ML20135J139
| ML20135J139 | |
| Person / Time | |
|---|---|
| Site: | 07003027 |
| Issue date: | 08/01/1985 |
| From: | PUBLIC SERVICE CO. OF NEW HAMPSHIRE |
| To: | |
| Shared Package | |
| ML20135J134 | List: |
| References | |
| NUDOCS 8509250307 | |
| Download: ML20135J139 (135) | |
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SPECIdl NUCiFAR MATERf4L LICENSE APPLICATION
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NEWH5MPSHIRE'YANKFF SEABROOKSTATIONUNIT1 l$E'*IOSEE$7%*
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T TABLE OF CONTENTS 1.0 GENERAL INFORMATION 1.1 Reactor and Fuel 1.1.1 Site Description 1.1.2 Owners and Constructors 1.1.3 Reactor Description 1.1'. 4 Fuel Assembly and Appurtenances Description 1.1.5 Nuclear Fuel 1.2 Storage Conditions 1.2.1 Design Bases 1.2.2 Fuel Storage Facilities 1.2.3 Fuel Handling 1.2.4 Fuel Handling System Design Bases 1.2.5 Safety of Storage and Adjacent Area Activities 1.2.6 Fire Alarm and Fire Control Systems 1.3 Physical Protection 1.4 Transfer of Special Nuclear Material 1.4.1 Transfer 1.4.2 Special Nuclear Material Control and Accounting Practices-1.5 Financial Protection and Indemnity 2.0 RADIATION CONTROL 2.1 Health Physics 2.1.1 Training and Experience of Health Physics Personnel 2.1.2 Contamination Detection Procedures 2.1.3 Contamination / Radiation Detection Equipment 2.1.4 Personnel Control and Radiation Monitoring 2.1.5 Radioactive Waste 2.2-Nuclear Criticality Safety 2.2.1 Qualification of Personnel 2.2.2 Responsibilities of Key Personnel 2.2.3 Storage of Fuel Elements 2.2.4 Nuclear Safety Analysis i
2.2.5 Enrichment Used for Criticality Calculation 2.2.6 Neutron Absorber Materials in Spent Fuel Pool Racks 2.2.7 Moderator control 2.2.8 Method Verification 2.2.9 Fuel Removal from Storage 2.3 Accident Analysis 2.3.1 Exemption From 10CFR70.24 Requirements 3.0 -
OTHER MATERIALS REQUIRING NRC LICENSE 3.1 Irradiation Test Capsules 3.2 Material of Any-Form
- 3.3 Incore Fission Detectors 3.4 Ex-Core Detectors 3.5 Storage and control of Material
I 1.0 GENERAL INFORMATION This application for Seabrook Station Unit 1 is being filed, pursuant to 10CFR70.21, on behalf of the Joint Owners (Subsection 1.1.2) by their agent, the New Hampshire Yankee Division (NHYD) of Public Service of New Hampshire to which construction and operating responsibilities have been delegated
(
Reference:
" Amendment 53 to March 30, 1973, Application to Construct and Operate Seabrook. Station Unit 1 and Unit 2", dated September 25, 1984).
In particular this is an application for a license authorizing receipt, possession and storage of unieradiated special nuclear material, which is described in Subsection 1.1.5, and possession, storage and use of associated radioactive material described in Section 3 at Seabrook Station Unit 1.
Specific details regarding fuel, storage condition, transfer of special nuclear material, nuclear criticality safety, etc. are provided hereinaf ter.
1.1 Reactor and Puel-1.1.1 Site Description The site property consists of approximately 896 acres near the northern boundary of the town of Seabrook, Rockingham County, New Hampshire. The site is situated about eight miles southeast of the county seat of Exeter;.five miles northeast of Amesbury, Massachusetts; and two miles west of Hampton Harbor inlet. The site is bordered on the east by an extensive saltwater marsh and is located on a point of land called "The Rocks", between two small tidal estuaries, the Brown's River.and the Hunt's Island Creek. The center of the Boston metropolitan area is approximately 40 miles south-southwest of the site. Figure 1.1-1 shows the site location in relation to principal cities and towns within a 5-mile radius.
4-Geographical coordinates of each reactor unit are as follows:
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Universal Transverse Latitude and Lonnitude Mercator Coordinates Unit No. 1 N 42 53' 55.4" 4751005 m N (Zone 19) i W 70 50' 58.7" 348994 m E Unit No. 2 N 42 53' 52.67" 4750928 m N (Zone 19)
W 70 51' 04.2" 348862 m E l-1 The site boundary will be the exclusion area, as defined in 10CFR Part 100.
The minimum exclusion radius is 3000 feet measured from the center of either containment building to the-nearest property line. All areas within the site boundary, with the exception of the railroad easement, the underground power line easement,'and portions of the Brown's River and Hunt's Island Creek are owned in common by the Seabrook joint owners.
i There are no industrial or recreational facilities, or residential homes located within the site boundary. However, New Hampshire Yankee operates a
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public information center on-site approximately 1200 feet southwest of Unit No. 2.
1.1.2 Owners and constructors The ownership of the Seabrook Project was delineated in the General and Financial portion of the Operating License Application for Seabrook Station Units 1 and 2 (
Reference:
" Amendment 43 to March 30, 1973, Application to
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Construct and Operate Seabrook Station Unit 1 and Unit 2; Submittal of j
Seabrook Station Final Safety Analysis Report, Seabrook Station Environmental Report - Operating License Stage, and Seabrook Station General and Financial Information", dated October 1, 1981). Currently, the composition and
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ownership;of the Joint Owners of Seabrook Station is as follows:
Public' Service Company of New Hampshire 35.56942%
The United Illuminating Company 17.50000 Massachusetts Municipal Wholesale Electric Company 11.59340 6 _
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New England Power Company 9.95766 Central Maine Power Company 6.04178 The Connecticut Light and Power Company 4.05985 Canal Electric Company 3.52317 Montaup Electric Company 2.89989 Bangor Hydro-Electric Company 2.17391 New Hampshire Electric Cooperative, Inc.
2.17391 Central Vermont Public Service Corporation 1.59096 Maine Public Service Company 1.46056 Fitchburg Ges and Electric Light Company 0.86519 Vermont Electric Generation and Transmission Cooperative 0.41259 Taunton Municipal Lighting Plant 0.10034 Hudson Light and Power Department 0.07737 100.00000%
PSNH has contracted with Yankee Atomic Electric Company (YAEC) of Framingham, Massachusetts for the services of certain personnel involved in project engineering, licensing, and fuel supply.
In addition PSNH has contracted with YAEC for establishing and implementing the Quality Assurance Program for design and construction. YARC will also provide engineering services necessary to support the operation of Seabrook Station.
Westinghouse Electric Corporation has been contracted to design, fabricate, i
and deliver the Nuclear Steam Supply System (NSSS) and nuclear fuel for the plant. Westinghouse will also provide technical assistance for installation and startup of their supplied equipment.
United Engineers & Constructors (Philadelphia, Pennsylvania) has been contracted to provide engineering, design, and procurement services for the balance of plant as well as to act as the constructor.
General Electric has been contracted to provide and install the turbine generators and associated equipment. General Electric will also provide technical assistance for'startup of their supplied equipnent.
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1.1.3 Reactor Description Seabrook Station Unit 1 is a four-loop, pressurized water reacter with support auxiliary systems designed by Westinghouse Electric Company. This unit is similar in design to Duke Power Company's W.B. McGuire Nuclear Station Texas Utility Generating Company's Comanche Peak Station, and Commonwealth Edison Company's Byron and Braidwood Nuclear Plants.
The reactor for Unit 1 is housed in a steel-lined reinforced concrete containment structure and a concrete containment enclosure structure. These structures were designed by United Engineers and Constructors Inc., who is also responsible for the balance of plant design.
l Unit I has been granted Construction Permit CPPR 135 and Docket Number 50-443.
Each unit will be initially operated at core levels up to and including 3,411 megawatts thermal which corresponds to a nuclear steam supply system thermal output of 3,425 megawatts thermal and a corresponding gross l
electrical output of 1,198 megawatts electric. The engineered safety features have been evaluated at a rating of 3579 megawatts. thermal which is the ' design rating.
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1.1.4 Fuel Assembly and Appurtenances Description Two hundred and sixty-four fuel rods are mechanically joined in a square, 1
17 x 17 array to form a fuel assembly as shown in Figure 1.1.4-1.
The fuel rods are supported at intervals along their length by grid assemblies which maintain the lateral spacing between the rods throughout the design life of the assembly. The grid assembly censists of an " egg-crate" arrangement of interlocked straps. The straps contain spring fingers and dimples'for fuel rod support as well as coolant mixing vanes. The fuel rods consist of slightly enriched uranium dioxide ceramic cylindrical pellets contained in slightly cold worked Zircaloy-4 tubing which is plugged and seal-welded at the ends to encapsulate the fuel. All fuel rods are pressurized with helium during fabrication to reduce stresses and strains and to increase fatigue life.. _ _ _ _ _ _ _ - _ _ _
The center position in the assembly is reserved for the incore instrumentation, while twenty-four (24) other positions in the array are equipped with guide thimbles joined to the grids and the top and bottom nozzles. Figure 1.1.4-2 shows the positions of all guide thimbles. Depending upon the position of the assembly in the core, the guide thimbles are used as core locations for Rod Cluster Control Assemblies (RCCAs), neutron source assemblies, and burnable poison rods. Otherwise, the guide thimbles are fitted with thimble plugs to limit bypass flow.
The bottom nozzle is a box-like structure which serves as a bottom structural element ~of the fuel assembly and directs the coolant flow to the assembly.
I The top nozzle assembly functions as the upper structural element of the fuel assembly, in addition to providing a partial protective housing for the RCCA I
and other components.
The fuel assemblies are designed to withstand loads induced during shipping and handling. All fuel assemblies are checked for enrichment and dimensions prior to final release from manufacturer. Table 1.1.4-1 gives materials of construction and pellet information.
Appurtenances used with the fuel assemblies include poison assemblies, primary and secondary sources, thimble plug assemblies and rod cluster control assemblies. A description of these appurtenances follows:
A.
Poison Assemblies i
Each burnable poison assembly consists of burnable poison rods attached to a holddown assembly. A burnable poison rod is shown in Figure 1.1.4-3.
When needed for nuclear considerations, burnable poison rods are inserted into selected thimbles within fuel assemblies.
The poison rods consist of borosilicate glass tubes contained within Type 304 stainless steel tubular cladding Which is plugged and seal-welded at the ends to encapsulate the glass. The glass is also supported along the length of its inside diameter by a thin wall tubular inner liner. The top end of the liner is open to permit the diffused helium to pass into the. - -.
void volume, and the liner. overhangs the glass. The liner has an outward flange at the bottom end to maintain the position of the liner with the i
glass.
1 The poison rods in each fuel assembly are grouped and attached together at l
the top and of the rods.to a holddown assembly by a flat, perforated retaining plate which fits within the fuel assembly top nozzle and rests 1
on the adaptor plate.
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The cladding of the burnable poison rods is slightly cold-worked Type 304 stainless steel. All other structural materials of the poison rods is Type 304 or 308 stainless steel except for the springs which are Inconel-718.
B.
primary and Secondary Sources 1
Primacy and secondary source rods have the same cladding material as the control rods in the Rod Cluster Control Assembly.. The secondary source rods contain pellets (Antimony-Beryllium) stacked to a height of approximately 88 inches. The primary source rods contain' capsules of Californium source material and alumina spacer to position the source I
material within the cladding. The rods in each assembly are permanently fastened at the top end to a holddown assembly.
The other structural members such as the spider head and vanes are l
constructed of Type 304 stainless steel except for the springs. The springs exposed to the reactor coolant are Inconel-718.
Figures 1.1.4-4 and 1.1.4-5 show primary and secondary source rods, respectively, i
C.
Thimble plum Assemblies i
L Thimble plug assemblies limit bypass flow through the guide thimbles in l
fuel assemblies which do not contain control rods, source rods, or
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burnable poison rods.
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The thimble plus assemblies consist of a flat base plate with shoc rods suspended from the bottom surface and a spring pack assembly as shown in Figure 1.1.4-6.
The 24 short rods, called thimble plugs, project into the upper ends of the guide thimbles to reduce the bypass flow. Each thimble plug is permanently attached to the base plate by a nut which is lock-welded to the threaded end of the plug. Similar short rods are also used on the source assemblies and burnable poison assemblies to plug the ends of all vacant fuel assembly guide thimbles.
i All components in.the thimble plug assembly, except for the springs, are constructed from Type 304 stainless steel. The springs are Inconel-718.
D.
Rod Cluster Control Assembly The silver-indium-cadmium (Ag-In-Cd) rod cluster control assembly is comprised of 24 neutron absorber rods fastened at the top end to a common spider assembly, as illustrated in Figure 1.1.4-7.
The absorber material used in the control rods are silver-indium-cadmium alloy slugs which are essentially " black" to thermal neutrons. As-In-Cd slugs are sealed in cold-worked Type 304 stainless steel tubes to prevent the absorber materials from coming in direct contact with the coolant.
The bottom plugs are made bullet-nosed to reduce the hydraulic drag during reactor trip and to guide smoothly into the dashpot section of the fuel assembly guide thimbles.
The absorber rod end plugs are Type 308 stainless steel. The design stresses used for the Type 308 material are the same as those defined in the ASME Code,Section III, for Type 304 stainless steel. At room temperature the yield and ultimate stresses per ASTM-580 are the same for the two alloys.
In view of the similarity of the alloy composition, the temperature dependence of strength for the two materials is also assumed to be the same.
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i The spider assembly'is in the form of a central hub with radial vanes
- containing cylindrical fingers from which the absorber rods are suspended. Detents for connection to the drive rod assembly and handling a
detents are machined into the upper end of the hub. A coil spring inside the spider body absorbs the impact energy at the end of a trip insertion.
l The radial vanes are joined to the hub by tack. weld and brazing. and the fingers are joined to the vanes by brazing. A center-post which holds the spring and its retainer is threaded into the hub within the-skirt and welded to prevent loosening in service. All components of the spider assembly are made from Types 304'and 308 stainless steel except for the retainer which is of 17-4 PH material and the springs which are Inconel-718 alloy..
The absorber rods are fastened securely to the spider. The rods are first j
threaded into the spider fingers and then pinned to maintain joint tightness, after which the pins are welded in place. The end plus below j
the pin position is designed with a reduced section to permit flexing of j
the rods to correct for small misalignments.
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1.1.5 Nuclear Fuel l-Special nuclear n=terial will be contained in a number of fuel assemblies which will not exceed one hundred ninety-nine (199) although only one hundred ninety-three (193) fuel assemblies are required to load the reactor core of l
Seabrook Station Unit 1.
Authorization for storage of 199 assemblies is i'
requested to allow for contingencies. The maximum enrichment of uranium in i
Uranium-235 is 3.5% by weight.
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The total weight of enriched uranium is about 461 Kg (U) per fuel assembly.
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The total weight of a fuel-assembly is 1,467 pounds.
l Seabrook Station will store not more than 92,000 Kg of enriched. uranium (3.5%
or less U-235).
No depleted uranium and only trace quantities of U-233, Pu, 1
l and Th are contained in the enriched uranium. Total U-235 content of the enriched uranium is 3,220 Kg or less.
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l 1.2 Storate conditions 1.2.1 Desian Bases i
i The fuel storage facilities are located within the Fuel Storage Building and are. designed to facilitate the safe handling, inspection, and storage of new -
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fuel assemblies and control rods.
The Fuel Storage Building, Figure 1.2.1-1, is a Seismic Category I building with an operating floor 5 feet above grade. The fuel storage area is protected against external tornado missiles by 2-foot thick reinforced concrete walls. The large roll-up door on the west wall of the Fuel Storage
.Bu ld ng is not designed for tornado missiles; however, a missile wall is i i l
provided inside the building to prevent any missiles that could possibly j
penetrate the roll-up door from reaching the storage pool or cooling equipment, t
f Other Seismic Category I equipment includes the spent fuel storage racks, I
spent fuel pool, fuel transfer tube, fuel transfer flange, and fuel transfer J-J outer sleeve.
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Seismic Category 1 equipment is designed to withstand the forces of the j
Operating Basis Earthquake (OBE) and Safe Shutdown Earthquake (SSE). For normal conditions plus OBE loadings, the resulting stresses are limited to j.
allowable working stresses as defined in the ASME Code,Section III, Appendix
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IVII, Subarticle XVII-2200 for normal and upset conditions. For normal
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conditions plus SSE loadings, the stresses are limited to within the allowable I
values given by Subarticle XVII-2110 for critical parts of the equipment which I
are required to maintain the capability of the equipment to perform its safety j
function. permanent deformation is allowed for the loading combination which I
includes the SSE to the extent that there is no loss of safety function.
For non-nuclear safety equipment, design for the SSR is considered if failure.
might adversely affect a Safety Class 1, 2, or 3 component. Design for the 4
OBE is considered if failure of the non-nuclear safety component might
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i adversely affect a Safety class 1 or 2 component.
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1.2.2 Fuel Storate Facilities The designated storage area for new fuel still in shipping containers is inside the Fuel Storage Building in the rail bay and the new fuel shipping container area (see Reference 1, FSAR Figures 1.2-16 and 1.2-18).
During this storage there are no criticality concerns because the shipping containers will be-stored in the configuration permitted per the requirements of the U. S. NRC Certificate of Compliance.(No. 5450) for that container. This certificate of Compliance allows the storage of up to 60 containers in any array without separation distance since analysis shows there is no criticality problem.
In the unlikely event that more containers are received than can be stored in the rail bay and new fuel shipping container area, a temporary holding area outside and adjacent to the Fuel Storage Building will be set up.
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.The designated storage area for new fuel that has been received and inspected is the spent fuel pool. The new fuel storage vault and associated racks may i
be used during receipt for inspection of new fuel. The new fuel storage vault will not be used for storage until it meets Seismic Category I requirements.
A description of the new fuel storage vault is contained in Reference 1.
The dust cover for the new fuel storage vault consists of removable steel i
plates which cover the entire vault (see Reference 1 FSAR Figure 1.2-16).
I The dust cover for the assemblies stored in the spent fuel pool consists of plastic sheets placed on top of storage racks containing new fuel.
2 The design basis of the fuel storage facilities includes the following:
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The spent fuel pool storage facility is designed in accordance with Regulatory Guide 1.13.
o The spent fuel racks are designed for high density fuel storage, and contain neutron absorbing material to assure a K,gg 1 0.95, even if the fuel is immersed in unborated water. L
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l o- - The facility and the building in which it is housed is capable of withstanding the effects of extreme natural phenomena, such as the SSE, tornadoes, hurricanes, missiles, and floods.
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The spent fuel storage racks have been designed to withstand separate loadings of an SSg, impact, handling loads, and dead load of the fuel assemblies, and meet ANSI N18.2 requirements.
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The impact load for the design of the racks is based on a 17 x 17 fuel f
assembly, 8.426 inches square, 167 inches long, weighing 1,467 pounds, falling 18 inches to the spent fuel pool racks and 30 feet to the new fuel storage vault racks at'ths worst possible orientation, o
The pool walls, fuel storage racks, and other critical components whose failure could cause criticality or physical damage to fuel, are classified as Seismic Category I.
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o Failure of nonsafety-related systems or structures located in the vicinity of the spent fuel storage facility which are not designed to Seismic-Category.I requirements will not cause an increase in K,gg to exceed the l
maximum allowable.
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o The spent fuel pool bridge and hoist is designed to remain on its rails
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during an SSE and, therefore, cannot damage stored fuel.
o The crane handling system is designed to prevent excessive forces from being applied to the spent fuel storage racks.
1 1.2.3 Fuel Handlina J
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Upon arrival on-site of a fuel delivery, the truck and shipping containers will be surveyed by Health Physics for contamination and inspected by Reactor Engineering for damage. When the truck is released, it will be sent to the Fuel Storage Building where the shipping containers will then be off-loaded and placed inside the building. Upon opening the shipping' containers, the internals will be inspected for damage and surveyed for contamination. After each assembly has been removed from a container it will be surveyed for i.
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contamination and inspected. The inspection takes place in the new fuel
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storage vault and includes visual checks for bowing, twisting, foreign debris,
- cracks, scratches, pits and burrs and dimensional checks of nozzles, fuel rod spacing and thimble tubes. At this time other core components such as control rods will undergo a similar inspection. When the inspection is complete, the assembib s will be transferred to the spent fuel pool for storage until core load.
1.2.4 Fuel Handling System Design Bases The primary design requirement of the fuel handling equipment is reliability.
A conservative design approach is used for all load-bearing parts. Where possible, components are used that have a proven record of reliable service.
f The following design bases were used for fuel handling system equipment assemblies during transfer of fuel assemblies:
o Fuel handling devices have provisions to avoid dropping or jamming of fuel assemblies during. transfer operation.
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Handling equipment used to raise and lower spent fuel has a limited maximum lift height.
o The Fuel Transfer System (FTS) has provisions to preserve the integrity of the containment pressure boundary. Figure 1.2.3-1 shows the Fuel Transfer
- System, o
Handling equipment will not fail in such a manner as to damage Seismic Category I equipment or fuel in the event of a SSE.
o The static design load for the refueling machine crane structure and all its lifting components is normal, dead and live loads, plus three times the fuel assembly weight with a rod cluster control assembly.
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o The allowable stresses for the refueling machine structures and components supporting a fuel assembly are as specified in the ASME Code,Section III, Appendix XVII, Subarticle XVII-2200.
Allowable stress criteria for rated loads for the spent fuel pool bridge and hoist, cask handling crane, and polar gantry crane are in accordance with CMAA-70.
o A single finger on the fuel gripper can support the weight of a fuel assembly and rod cluster control assembly without exceeding the designed allowable stresses for the refueling machine structures and components, o
The design loading on wire rope hoisting cables does not exceed 0.20 times the average breaking strength.
Two cables are used in the refueling machine and each is assumed to carry one half the load, o
All components critical to the operation of the equipment are located so that parts which can fall into the reactor are assembled with the fasteners positively restrained from loosening under vibration, The inertial loads imparted to the fuel assemblies or core components o
during handling operations are less than the loads which could cause damage.
Physical safety features are provided for personnel operating handling o
equipment.
Industrial codes and standards used in the design of the fuel handling equipment:
o Refueling Machine Design: Appilcable sections of CMAA Specification No. 70.
o New Fuel Elevator Holst: Applicable Sections of HMI-100 and ANSI B30.16.
o Structural: ASME Code,Section III Appendix XVII, Subarticle XVII-2200 (Refueling Machine). J
o Electrical: Applicable standards and requirements of the National Electric Code and NFPA No. 70 are used in the design, installation, and manufacturing of all electrical equipment.
o Materials: Main load-bearing materials conform to the specifications of the various ASTM Standards.
o Safety: OSHA Standards 29CFR1910 and 20CFR1926 including load testing requirements; the requirements of ANSI N18.2, Regulatory Guide 1.29 and General Design Criteria 61 and 62.
Additional details on the refueling machine, spent fuel poc1 bridge and hoist, cask handling crane, new fuel elevator, fuel transfer system, rod cluster control changing fixture, spent fuel assembly handling tool and new fuel assembly handling tool are contained in Reference 2.
1.2.5 Sa'fety of storate and Adjacent Area Activities Approved operating procedures control all fuel and heavy equipment movement.
Heavy equipment movement procedures follow the guidance provided in NUREG-0612. " Control of Heavy Louds at Nuclear Power Plants", Section 5.1.2, Spent Fuel Pool Area - PWR.
Fuel movement from storage is necessary in order to perform the tasks of inspection, procharacterization, and indexing.
o Inspection:
This involves a visual or otherwise nondestructive examination of the fuel assembly to determine its acceptability before exposure in the reactor core.
o Precharacterization: This involves measurements and examination to determine the pre-exposure characteristics of a given fuel assembly.
o Indexing: One assembly may be removed from storage and inserted into various locations of the reactor vessel for purposes of indexing the core.
All sections of the fuel storage facility which require that heavy loads be moved over the fuel storage area while this license is in effect will be substantially complete prior to receipt of fuel.
Fuel handling equipment..
shall be complete and preoperationally tested prior to moving fuel with the equipment. There will be no immediately adjacent area activity which may affect fuel storage safety.
To assure that the design of the spent fuel bridge / hoist is adequate to withstand the SSE without loss of structural integrity, the following measures have been implemented:
o A seismic analysis has been performed (by the vendor) in accordance with the design control procedures for Seismic Category I components for this crane to demonstrate that it will not lose its structural integrity during or subsequent to an OBE or SSE.
o Quality control requirements for material processing, welding, and nondestructive examination have been incorporated into the relevant specifications and purchase order documents for this crane to provide a level of confidence in the material and fabrication procedures being used to assure that the intent of Section C2 of Regulatory Guide 1.29 has been met.
The crane vendor shall be required to provide Certificates of Compliance o
for all load carrying members used in the fabrication of the crane.
The welding procedures used to fabricate the crane have been reviewed by o
UE&c.
o The spent fuel pool bridge and hoist is designed to remain on its rails during a SFP and, therefore, cannot damage stored fuel.
The cask handling crane is not a seismic Category I component; however, in compliance with Regulatory Guide 1.29 the crane design parameters are specified to provide adequate quality control of fabrication and design so that in the event of an Operating Basis Earthquake (OBE) or Safe Shutdown Earthquake (SSE), the crane will not fail in such a manner as to impair the functioning of any plant feature designated as seismic Category 1.
The crane is prevented from being dislodged off its rails during the SSE by mechanical anti-derailing devices.
1.2.6 Fire Alarm and Fire Control Systems the Fuel Storage Building is one large fire area and has been provided with an ionization fire detection system, manual firefighting equipment consisting of portable fire extinguishers, halon and dry chemical along with a standpipe and hose reel system. The fire loading is very low. The building is locked with controlled access. Administrative controls will control ignition source work and the storage and use of combustibles in the building. The Seabrook Station Fire Protection Program, Evaluation and Comparison to BTP APCSB 9.5-1 Appendix A Revision 2 Section F.1. Tab 9 has the fire hazard analysis for this building. The fire suppression, capability of the building was evaluated in Section 9.5.1 of the SER and found acceptable.
1.3 Physical Protection A description of the Seabrook Station Physical Security Plan for the Protection of Nuclear Material of Low Strategic Significance will be provided as a separate part of the application withheld from public disclosure. This plan was prepared pursuant to the requirements of 10CFR70.67, Physical Protection of Special Nuclear Material of Moderate and Low Strategic Significance.
1.4 Transfer of Special Nuclear Material 1.4.1 Transfer Westinghouse Electric Corporation of Columbia, South Carolina, fabricator of the nuclear fuel assemblies, is responsible for shipment of the fuel to Seabrook Station. All fuel assemblies are to be delivered to the site in accordance with shipping procedures and arrangements of the Westinghouse Electric Corporation authorized for use by that company under separate nuclear matorial license SNM-1107. The shipping container that will be used to ship the initial core to the site is supplied by Westinghouse and is covered by U. S. KRC Certificate of Compliance 5450.
Shipping quantities will be in accordance with the Certificate of compliance. _. _
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i 1.4.2 Special Nuclear Material Control and Accounting Practices I
The nuclear materials involved in this facility are uranium and plutcnium of f
various isotopic content. The Nuclear Materials Control System involves the j
establistument of: organizational lines for the safeguard of nuclear 2
materials, item control areas, recording and reporting procedures, receiving i
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and shipping procedures, internal transfer procedures, inventory procedures, 4
l element and isotopic calculation methods, and internal audit procedures, i
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The President of Yankee Atomic Electric Company's Nuclear Services Division i
j (YAEC-NSD) shall appoint within YAEC-NSD a Nuclear Materials Safeguards j
Manager. The Nuclear Materials Safeguards Manager shall have overall j
responsibility for the control and safeguard of special nuclear material.
In j
this function, he will be assisted by the Fuel Management Department Director and Reactor physics Manager, as well as others that may be delegated from Yankee Atomic Electric Company's Nuclear Services Division Engineering staff.
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The Station Manager has the ultimate responsibility for the control and
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surveillance of all SNM on the plant site. The Station Manager is kept j
informed about major operations involving SNM by a Nuclear Materials Custodian f
or his designated alternate.
The Reactor Engineering Supervisor at the station is responsible for ensuring
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qualified personnel perform SNM operations at the plant site. These operations include transferring and shipping of SNM at~the facility. During i
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any of these functions, personnel must follow applicable procedures.
i New fuel assemblies normally arrive via truck in containers designed for new fuel shipping. A packing list or o111 of lading, containing a listing of all I
the shipping containers in the shipment and an SNM distribution sheet, is l
included. Upon receipt, the serial numbers of shipping containers shall be checked against the packing list or bill of lading. After the off-loading and opening of shipping containers, fuel assembly serial numbers shall be verified l
against the shipping documents. Then fuel assembly history cards are made and I
placed in the SNM File. After storage, an inventory of the Item Control Area t
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(ICA) is completed and retained in the SNM File. This SNM File is maintained by the Reactor Engineering Department (RED) at the plant.
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The RED notifies the Central Accountability Office at YAEC-NSD of the time and date of arrival, assembly, serial numbers, storage location, and quantity of SNK contained in each shipment. All pertinent requisition, receipt, 4
inventory, disposal, transfer, import or export records shall be maintained for each assembly while on-site and for at least five years af ter transfer or export.
Fuel assembly movement is made in accordance with a Material Transfer Form (NTF) prepared and authorized by the RED.
Completed NTFs are returned to RED to update the fuel assembly history cards; Semi-annual inventories are performed on each ICA.
Losses upon verification are reported to the appropriate NRC Region Office. Completed inventories are 1
forwarded to the Central Accountability Office, i
A Material Status Report (Form DOE /NRC 742) shall be filed semi-annually.
The results contained therein shall reflect the most recent information available in the records. The report shall be prepared in accordance with the instructions, including the filing of a separate report for each type of SNM.
1.5 Financial Protection and Indemnity k
l The Joint Owners shall have obtained 1 million dollars of Nuclear Liability Insurance for Seabrook Station Unit 1 prior to receipt of fuel. This r
insurance coverage satisfies the requirements specified in the Price-Anderson Act for licensed nuclear power plants. Documentation of insurance will be i
available for inspection at the site prior to receipt of fuel.
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2.0 RADIATION CONTROL 2.1 Health Physics The Health Physics Department of Seabrook Station has been organized and equipped to protect plant employees from unnecessary exposure or contamination due to radiation and radioactive materials. Plant controls governing health physics and fuel handling will be reviewed by the Station Operation Review Committee (SORC) and approved by the Station Manager. Details of SORC are contained in Reference 3.
Personnel having health physics responsibilities meet the requirements of Regulatory Guide 1.8 (1977 Edition), except that ANSI /ANS 3.1-1978 will be used as the standard rather than ANS 3.1/ ANSI N18.1-1971 (Reference 4).
With regard to Section 4 of ANS 3.1-1978, individuals who do not possess the formal requirements specified in this section shall not be automatically eliminated where other factors provide sufficient demonstrations of their abilities.
(For further clarification and alternatives see Reference 4.)
2.1.1 Training and Experience of Health Physics Personnel The Health Physics Department Supervisor, who meets or exceeds applicable qualifications of ANSI 3.1, 1978, paragraph 4.4.4, is responsible for administering the station radiation protection program, and is responsible to the Station Manager for compliance with applicable federal and state radiation protection regulations. Appendix A contains his experience and training.
The Seabrook Station Health Physics Department Supervisor is the individual designated to be the Radiation Protection Manager as that position is described in ANS 3.1-1978 and Regulatory Guide 1.8, Revision 1-R.
His duties and responsibilities are outlined in FSAR Section 13.1.2.2.c.4.
His qualifications, as stated in FSAR Section 13.1.3.1, and as shown by his resume provided in Appendix A, equal or exceed those required in the above noted references.,
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The Health Physics Supervisors report to the Health Physics Department Supervisor and are responsible for supervising the routine and daily department operations, and providing assistance to the Department Supervisor.
The Health Physics Supervisors meet or exceed the minimum qualifications ac specified in ANSI 3.1, 1978, paragraph 4.3.2.
One of the Health Physics t
Supervisors may temporarily assume the responsibilities of Department Supervisor, when the Department Supervisor is absent for an extended period of time. Appendix A contains their experience and training.
s Health Physicists provide administrative and technical assistance to the
. Health Physics Department Supervisor'and Health Physics Supervisors. Their principal responsibilities include administration of the Seabrook Station Radiological Environmental Surveillance Programs, ALARA Program, and Personnel Dosimetry Program. Minimum qualifications are as specified in ANSI 3.1, 1978, paragraph 4.3.2.
The. Health Physics working foremen are responsible for the routine Health Physics activities and coordinating assignments of the Health Physics Technicians. Qualifications of the working foremen have been set in accordance with ANSI 3.1, 1978, paragraph 4.3.2.
3 Department Technicians are responsible for performing the routine and daily
{
operations of the department. Technicians meet at least the minimum qualifications applicable to their work, as specified in ANSI 3.1, 1978,
'I paragraph 4.5.2, and any other qualifications set by management to insure that the technicians will be capable of performing their duties efficiently and expertly. These duties include performing various surveys, collecting air i
samples, maintaining department equipment and instrumentation, and providing i
j radiation protection / control coverage, as necessary.
2.1.2 Contamination Detection Procedures Specific. limits associated with contamination control, sealed source leakage and actions to be taken are delineated in the Station Byproduct Material License No. 28-20818-01. See Reference 5 for applicable section of license l
application.
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Site receipt of licensed source and specisi nuclear material is performed in accordance with station procedures. These contamination detection procedures are implemented in the manner described below, preliminary radiation and contamination surveys are conducted on a new fuel shipment prior to protected area access to evaluate its radiological status.
Upon arrival on-site and pursuant with 10CFR20 and 49CFR173, a comprehensive survey is performed on the transport vehicle and the exterior of the shipping containers. Radiation surveys are performed using a portable Geiger-Mueller 1
survey instrument. Contamination levels are evaluated by performing a smear survey. Smears are counted for alpha and beta-sanuna contamination.
Radiological conditions exceeding the limits specified in 10CFR20.205 shall result in the establishment of appropriate radiological controls and notification of the NRC/ DOT.
During the opening of shipping containers, appropriate radiological precautions are implemented. Radiation and contamination surveys are performed on the interior of the container and container contents. periodic airborne radioactivity sampling is established during shipping container opening. As the fuel assembly protective covering is removed, smears and contact radiation readings are taken at representative locations.
All fuel handling activities are performed under a Radiation Work permit (RWP). The RWP specifies requirements for personnel dosimetry, protective clothing, and health physics work coverage responsibilities.
The fuel handling area is posted with appropriate caution signs and boundarles as radiological conditions warrant. A contamination frisking station is provided at the exit of the area during receipt, inspection, and movement of new fuel. A portable area radiation monitor with an audible alarm is located in the vicinity of the fuel handling area during fuel movement.
If contamination levels exceeding the site contamination limits are discovered, contamination control measures will be implemented. These measures will include establishing a contamination control point, strict control of personnel access to the area, use of appropriate protective __
l clothing, and re-establishment of area barriers, boundaries, and postings. As a precaution, periodic air sampling may be performed.
Decontamination of the area and equipment is initiated under the supervision of Health Physics personnel. Waste generated by the decontamination activities is handled under the requirements set forth by station procedures.
2.1.3 Contamination / Radiation Detection Equipment The station has a counting room that is equipped with radLation detection equipment to analyze routine contamination survey smears.
This equipment is capable of detecting alpha, beta, and samma activities at levels less than contamination control levels. This counting room equipment is used for quantitative and qualitative analysis of smear samples.
Portable radiation detection equipment consists of low-and high-range ion chamber dose rate meters, Geiger-Mueller count rate meters, scintL11ation alpha counters, neutron rate meters, and air samplers.
Sufficient quantities of each type of instrument are available to permit calibration, maintenance, repair, and handling of peak loads without diminishing the radiation protection program.
Information about instrumentation is presented in Table 2.1.1.
Equipment required to support fuel handling operations is normally stored at the fuel building control point where it is easily accessible.
Extra equipment not intended for daily use is stored in a Health Physics storage area.
Personnel external contamination detection equipment consists of hand held friskers with a range of 0 to 5 x 10 cpm. This equipment is designed for routine use by all personnel that exit a Radiologically Controlled Area (RCA).
Bionssay services are provided whenever an internal intako of radloactive material is known or suspected to have occurred. The Seabrook Health Physics Department may provido in vivo (whole body) counting elthor on-site or through -
contracted services.
If required, in vitro bionssay services are provided through contractor services.
Radiation and contamination detection equipment calibration and functional testing is performed by the Health physics technicians periodically in accordance with applicable station procedures.
Functional testing and calibration are also performed following major maintenance, questionable accuracy, and as deemed necessary due to unusual circumstances such as possible damage. Calibration is performed using sources or methods traceable to the National Bureau of Standards.
Fixed radiation detection equipment calibration and operational equipment checks are performed. Radiation background and detection efficiency factors are checked prior to equipment use each day.
Calibration of this equipment is performed periodically and includes efficiency checks, operational checks, and, when applicable, a plateau check.
CallSaation of portable survey equipment in use will be done on a semi-annual basis or as necessary due to major maintenance, questionable accuracy, or possible damage. Prior to using survey equipment, operational checks are performed which include the checking of battery conditions and performing a response check.
Calibration techniques of air samplers will vary according to type. The air samplers will be calibrated cemi-annually, after maintenance that may affect operation, and when accuracy is questionable.
The calibration procedure for the low volume, high volume, and personnel air samplers will consist of adjusting the flow rate meter or determining a new flow rate factor.
Continuous Air Monitor (CAM) calibration is dependent upon type chosen, but will generally consist of flow rate adjustments, source checking of detectors, and checking filter paper speed and adjustment, if necesonry.
2.1.4 Personnel Control and Radiation Monitoring Individuato working in radiologically controlled areas or with radionuclides receive general employee training. General employee training consists of site familiarization, security, radiation protection, industrial safety, 10CFR19, _
emergency plan and quality assurance. Training Department personnel with support from the Health Physics Department and other station personnel have developed the program content and testing. This training meets or exceeds the requirements given by 10CFR19.12.
The Training Manager reports directly to the Assistant Station Manager and is responsible for development, implementation and administration of Non-Licensed Training including General Employee Training.(CET), Fire Brigade Training and Specialized Training.
Instructors for CET are provided by the Training Department. Documentation of training is maintained by the Training Department in accordance with station procedures.
I Dosimeters are issued to individuals requiring access to fuel handling operations and are worn when participating in such operations, as dirac*.ed by Health Physics personnel. TLDs and self-reading pocket dosimeters gro the i
principle dosimetry devices used.
Appropriate occupational exposure information la recorded and maintained. The exposure information is used to generate the various reports required by 10CFR19 and 10CFR20.
2.1.5 Radioactive Weste Radioactive waste is handled in a manner as to keep personnel exposure as low as reasonably achievable.
Storage of waste is at a location designated by the Health Physics Supervisors. Removal of waste from storage is prohibited without permission from the Health Physics Department personnel.
Seabrook Station does not intend to ship radioactive waste prior to plant operation. In the event it is necessary, radioactive waote will be shipped to a facility licensed by applicable governmental agencies. Surveys will be performed prior to shipment to insure compliance with NRC regulations, DOT regulations, and any disposal restrictions.
Documentation of radioactive waste disposal will consist of name, address, phone number, and permit or license numbers of consignor and consigneo. A qualitative assessment and quantitative analysis will be performed and.
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documented. Applicable shipping documents are completed and maintained on file at the station with a copy provided to the driver. All waste is appropriately labeled and special instructions are given to drivers of
" exclusive use" vehicles.
2.2 puclear Criticality Safety i
2.2.1 Qualification of personnel The recommendations of Regulatory Guide 1.8, " personnel Selection and i
Training," Revision 1-R, have been used as the basis for establishing minimum qualifications for all management, supervisory and professional-technical personnel in the Station organization, with the exception that ANSI /ANS 3.1-1978 will be used as the standard in lieu of ANS 3.1/ ANSI 18.1-1971 (Reference 4).
The education, training and experience requirements for operators, technicians and mechanics will equal or exceed the qualifications for the positions stated in ANS 3.1-1978 and Regulatory Guide 1.8.
Established company training programs include documented academic and on-the-job training plus comprehensive qualification examinations applicable to the skill level of the position assignment. Where desirable, off-site facilities may be used for specialized training. Records of the scope, general content and level of accomplishment for each person attending off-site training are retained at the Station.
2.2.2 Responsibilitios of Fey personnel The Nuclear Quality Manager at the station has overall responsibility for r
assuring that the Seabrook Operational Quality Assurance program is effectively implemented by all organizations performing work on safety-related systems and equipment at Seabrook Station. This individual has education, training and experience which equals or exceeds that of ANSI /ANS 3.1-1978, paragraph 4.4.5.,
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1 The Reactor Engineering Department Supervisor is responsible for fuel movement j
sequences, fuel accountability, fuel storage, and fuel inspection. This individual has education, training and experience which equals or exceeds that of ANSI /ANS 3.1-1978, Paragraph 4.4.1.
The Fuel Management Department of Yankee Atomic Electric Company, Nuclear Services Division provides the l
independent nuclear accountability function, maintains in-process nuclear material inventory, and provides assistance to the Reactor Engineering Department Supervisor, as requested.
1 2.2.3 Storate of Fuel Elements Fuel elements will be stored temporarily in the fuel receipt area in shipping containers. The shipping containers will be taken to the Fuel Storage Building and the fuel removed and stored in the spent fuel pool in a timely manner.
2.2.4 Nuclear Safety Analysis Fuel spacing is maintained by the fuel storage racks which are Seismic Category I equipment. The racks consist of individual vertical cells.
f Administrative limitations on placement of fuel in the racks is not necessary l
according to criticality calculations. Placement and design of racks is such as to preclude insertion of fuel in places other than the vertical cells.
A design description of the fuel storage racks is provided in Subsection 1.2.2.
Loading, shock, fire, and corrosion are not credible methods of l
achieving criticality since a moderator such as water would have to be present. This would constitute a double fault situation since the fuel storage condition is without moderation. Additionally, loading was analyzed as part of the seismic I category ofsign. The Station Fire Protection Program Manual describes the administrative controls and pemit system for the control of ignition sources and the combustible materials permit system. The racks are designed for impact loading of one fuel assembly (17 x 17), 8.426 inches square, 167 inches lonc,. weighing 1,467 pounds, and falling at the worst possible orientation, 18 inches to the spent fuel racks and 30 feet to the new 1
fuel rucks. Racks are constructed of steel and should not be affected by credihte corrosive agents.
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2.2.5 Rnrichment Used-for Criticality calculation
+
Mew fuel storage ~ vault criticality calculations were performed assuming up to 4.0 w/o U-235.
This analysis was done with 17 x 17, 0.496 inch pitch
. Westinghouse standard fuel. Realistic modeling using this enrichment shows the racks to be suberitical over the entire range of moderator density. The criticality calculation was performed at 3.1 w/o U-235 to determine the point of optimum moderation. Additional calculations were performed at-that point for 3.5 and 4.0 w/o U-235.
All optimum moderation calculations were below the NRC criteria of 0.98.
The spent fuel rack criticality calculation shows K,gg less than 0.95 for an
)
enrichment of 3.5 w/o U-235.
This analysis.was done with 17 x 17, 0.496' inch pitch Westinghouse standard fuel. Realistic modeling shows the racks to be 1
suberitical over the entire range of moderator density.
Refer to FSAR Figures 9.1-20 and 9.1-23 contained in Reference 1 for further information regarding rack K,gg.
2.2.6 Neutron Absorber Materials in Spent Puel Pool Racks The neutron absorbing material, Boraflex, consists of boron carbide particles held in place by a nonmetallic binder. The elemental composition by weight percent of the product is as follows:
i Silicon 18.9%
?
Oxygen 17.0%
Carbon 21.0%
i Hydrogen 2.1%
l Boron 41.0%
4 The material contains a minimum of 0.020 gas of isotopic B10 per square centimeter. This material, which is manufactured by Brand Industrial Services, Inc. (BISCO), has been tested and qualified for spent fuel storage i
l applications. Boron. carbide of the grade in the Boraflex will typically l
i contain 0.01 to 0.15 weight percent of soluble boron. Test results confirm the encapsulation function of the' silicone polymer matrix in preventing the 1 _..
_- _ _ -. _ _ _ _ _, _. - _ _.,, _,._ _ _.. _ _. _ __-. _ -. _ __.~.-.., - - _.. _.... _ _ _
leaching of the soluble species from the boron carbide. The soluble contaminants fn ths boron carbide are not considered for certified Boraflex.
Boraflex is inserted on all sides of fuel holding cells which are not adjacent to the spent fuel pool liner. For dimensional information refer to Table 2.2.6-1 and FSAR Figure 9.1-14 (Reference 1).
By design the poison material is encapsul'ated in stainless steel for structural support. The material is not sealed, since it is compatible with the environment. Poison verification holes are included in each cavity wall for visual inspection to assure the presence of poison during fabrication.
The presence and effectiveness of the poison material is assured since:
Controls were imposed during manufacture to ensure material installation; The poison material is encapsulated and not designed for removal; The poison material has been shown to be compatible with its environment.
Westinghouse provided the spent fuel racks, which included the encapsulated boraflex, to Seabrook. The manufacturer's (Westinghouse Nuclear Component Division) quality program complies with Westinghouse Water Reactor Division QA Plan WCAP-8370. QA Plan WCAP-8370 has been reviewed by the NRC staff.
The installation of the completed spent fuel racks and control of the racks after receipt at the site is the responsibility of the applicant although Westinghouse may provide criteria and procedures in regards to installation.
Fuel racks are Category 1 QA items; therefore, inspection and installation are performed in accordance with the Station QA Program.
2.2.7 Moderator control The calculations performed on the spent fuel pool considered fresh fuel at an enrichment of 3.5 w/o U-235.
The pool water was considered to be at a temperature of 68 F and contained no boron. Under these conditions, an infinite array of canisters with a 10.35-inch center-to-center spacing gave a K,gg below the NRC limit of 0.95. l
i The spent fuel rack criticality analysis was done with conservative neutronic assumptions and " worst case" mechanical dimensions. One of the conservative neutronic assumptions was unborated water at 68 F (0.9982 gm/cc). This l
density of water is optimum for the spent fuel rack criticality analysis. Dry
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conditions or conditions of foam or mist in the spent fue1 rack are less reactive than the optimum moderation. This is due to the flux trap principle and the close spacing between assemblies.
Therefore, the analysis with optimum moderation, boundh both the dry and foamy conditions.
In contrast, the new fuel vault is normally dry with wide spacing between assemblies for criticality control. Optimum moderation in the new fuel vault occurs when foam or mist is present in the wide space between assemblies'.
The calculations performed for the new fuel vault evaluated K,gg as a function of moderator density. Structural material in the new fuel racks acts as a neutron absorber but was not used in the calculations. Not including the structural material yields a conservative calculation. Water was assumed to be the moderator material.
K,g remains below 0.95 for all densities between and including 0 and 1.
This indicates there is no water density including fire-fighting mist, sprinkler system water, or flood, that would pose a problem.
The probability of water addition is very low. No sprinkler system exists in the fuel storage areas. Flooding probabilities were calculated for different areas of the site and are low for the fuel vault / pool.
Fire-fighting teams and personnel are directed to use dry chemical extinguishers. Additional fire information is supplied in Subsection 1.2.6 of this application.
Water retention around or in fuel bundles in the new fuel storage vault is not possible when the area is flooded and drained. Material covering individual bundles is removed prior to storage. Dust covers for new fuel are discussed in Subsection 1.2.2.
2.2.8 Method Verification Prior to use of any calculational tool in criticality calculations.
ANS 8.11/ ANSI N16.9-1975 standards require validation to assure applicability to the problems of interest. For a fuel storage rack calculation, our -
r validation consists of comparing KENO-IV results to critical experiments. The mechanical design of high density storage racks makes use of a flux trap principle using, for example, Boral poison sheets. Boral consists of a matrix of 35 w/o natural B C in a Type 1100 aluminum alloy matrix clad with Type 4
1100 aluminum alloy. Critical experiments which utilized some form of B C 4
were required to provide adequate validation.
Several critical experiments are now available which utilize B C in a 4
pseudo-fuel storage rack configuratton. These consist of the Bierman criticals from Battelle Pacific Northwest Laboratories and B&W criticals.
The Bierman criticals were performed using three suberitical clusters with and without poison plates inserted between clusters. Two different enrichments were utilized, 2.35 and 4.30 w/o.
In addition, Boral was used as one of the poison materials. The calculated values and their uncertainties for these experiments are shown in Table 2.2.8-1.
It should be noted that R. Westfall of ORNL has generated KENO IV comparisons using the 123-group library for all 210 of the Bierman criticals with no calculational bias. Comparison of individual cases calculated by Yankee to the ORNL results show agreement well within the uncertainty of the calculational method.
Tabic 2.2.8-2 provides the results of the Yankee calculation of the B&W critical experiments. The B&W critical experiments were conducted using low-enriched (2.46%) UO fuel pins in a water-moderated lattice of fuel assemblies,that simulated a variety of close-packed LWR fuel storage configurations. The experiments were performed at the B&W Lynchburg Research Center in the CX-10 critical facility. Three cases were analyzed, Core I, Core XVI and Core XIX. Core I is a cylindrical array of 438 fuel pins and represented the B&W base case. Core XVI contained nine fuel pin clusters grouped as a 3 x 3 array with Boral sheets. The Core III configuration is similar to Core XVI, how:nar, this array utilized a high soluble boron concentration (634 ppm) which is an applicable parameter in some criticality calculations. The resulis in both Tables 2.2.8-1 and 2.2.8-2 demonstrate the accuracy of the calculational methods.
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The MITAWL-KENO-IV method with the 123-group library has been validated against experiment and detailed Monte Carlo calculations for spent fuel rack geometries with poison. However, MITAWL-KENO-IV criticality calculations for i
new fuel storage with fire-fighting foam or mist have not been benchmarked against experiment or detailed calculations. Because of this, it was suggested that an independent calculation be performed with continuous energy Monte Carlo and modern cross section data. Brookhaven National Laboratory performed the same new fuel storage rack criticality calculations with the SAM /CE Monte Carlo code and modern ENDF/B-V cross section data. Their calculation represents the highest level of computer benchmarking short of an actual experiment.
a For a flooded (0% void) new fuel vault, the SAM /CE value of K,gg was 0.8546 l
1 0044 compared to the KENO value of 0.8587 1 0060 using a three-dimensional
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model of the new fuel vault. For 97.5% void, -the SAM /CE value of K,gg was 0.8090 1 0048 compared to a KENO value of.8414 1 0051. As can be seen, in these cases, the KENO calculation is more conservative than SAM /CE.
Documentation of verification is contained in Yankee Atomic Electric Company's Reports 1224 and 1343. Details of the calculations and assumptions used are contained in Yankee Atomic Electric Company's Report 1343.
2.2.9 Fuel Removal from Storate As discussed in more detail in Sections 1.2.3 and 1.2.4, fuel assemblies are removed from storage in the shipping cor.td ners and moved to.he new fuel p
storage vault / racks for inspection and ;?echa-t-terisation. hpon completion of these activities, the assemblies are moved to the spent fuel pool storage racks for storage until core load.
At no time can there be more than two (2) assemblies out of shipping containers or out of the new/ spent fuel racks because of crane 1' imitations.
These assemblies will.be maintained a safe distance from each other (i.e.,
cannot achieve a critical configuration) due to the location of the cranes and the crane travel.
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If, at-the time of fuel receipt, the new fuel racks are not seismically qualified using the current Amplified Response Spectra (ARS) not more than twelve (12) assemblies.shall be placed in the new fuel racks for purposes of inspection and characterization. The placement of assemblies into the new fuel racks shall be administratively controlled to preclude any criticality concerns (i.e., placement of fuel in racks in at least a checkerboard fashion). Administrative controls provide adequate measures to assure nuclear criticality safety because:
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A.
The new' fuel storage racks were seismically qualified using an ARS which is approximately 67% lower than the current ARS.
However, this qualification analysis was based on fully loaded racks (i.e., 90 assemblies). Therefore, limiting the number of assemblies in the fuel racks to 12 assemblies (approximately an 87% reduction) would assure the structural integrity of the racks. That is, as a minimum, it is not credible to consider gross failure of the racks due to a seismic event with the reduced fuel loading.
B.
The checkerboard spacing, which would be the ILmiting spacing condition used under administrative control, provides greater spacing between the fuel assemblies than the as-designed spacing used in the criticality analysis for the racks with 3.1% w/o U-235 (reference Section 2.2.5).
C.
The as-designed spacing is also conservative since under optimum conditions for moderation, the racks were shown to be suberitical with 4.0% w/o U-235.
D.
In order to obtain optimum conditions for moderation, a double fault must have occurred as discussed in Section 2.2.4.
E.
Approximately 2/3 of the fuel assemblies to be received contain either 2.4% or 1.6% U-235.
2.3 Accident Analysis A new fuel assembly accident in which the necessary geometry and moderator are provided for criticality is highly improbable. Subsections 1.2.2, 1.2.3, and l
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7-i 1.2.4 discuss design and administrative procedures to prevent accidents.
These procedures includo how to handle leakage of radioactive materials, and decontamination of personnel and equipment.
In the event physical damage does occur to new fuel stations, response would be in accordance with radiation protection procedures.
2.3.1 Exemption from 10CFR70.24 Requirements The applicant requests exemption from the monitoring and emergency procedure requirements of 10CFR70.24. This exemption request is based on special nuclear material storage arrangements and controls the applicant proposes to employ along with analytical calculations showing K,gg to be less than 0.95 for credible circumstances. These controls and storage conditions include:
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Limiting the number of assemblies out of storage as specified in Subsection 2.2.9.
o Storage conditions as specified in Subsection 1.2 to 1.2.6.
o Physical protection as specified in Subsection 1.3.
o Approvals as specified in Subsection 2.1.
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3.0 OTHER MATERIALS REQUIRING NRC LICENSE 3.1 Irradiation Test Capsules The Seabrook Station Unit I reactor vessel material surveillance program will utilize radioactive materials as dosimeters. Six test capsules will be provided for the reactor vessel, and will contain radioactive dosimeters (Uranium 238 and Neptunium 237) in.the following concentrations:
Uranium 238
~ microcuries (per capsule)
Approximately 4.0 x 10 Approximately 24.0 x 10~ microcuries (total per 6 capsules)
Neptunium 237 Approximately 12.1 microcuries-(per capsule)
Approximately 72.6 microcuries (total per 6 capsules)
The Uranium 238 (approximately 12 milligrams per capsule) will be supplied as U
Powder encapsulated in stainless steel capsules, and the Neptunium 38 237 (approximately 20 milligrams per capsule) will be supplied as Np0 powder encapsulated in stainless steel capsules (sealed sources). The capsules containing these radioactive materials are sealed in steel blocks which are prepared by Westinghouse and are inserted in each reactor vessel irradiation surveillance test capsule.
3.2 Material of Any Form-In addition to other special~ nuclear material, the Licensee will possess the following radionuclides in any chemical or physical form for use in calibration or chemical analysis:
Plutonium-238 (100 microcuries)
Plutonium-239 (100- microcuries).
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Uranium-235 (200 microcuries)
Uranium-238 (300 microcuries) 3.3 Incore Fission Detectors The Station will have 25 subminiature fission chambers. These sealed source fission detectors are Westinghouse Model Number WL-23957. Each detector contains 12.7 milligrams elementary uranium as U 03 8 enriched to 93% by weight U-235.
The total quantity is 12.7 *
(25 chambers) or 67.5 b
milligrams.
3.4 Ex-Core Detectors The Station will have 5 qualified excore detectors. These sealed-source excore detectors are Gamma Metrics Model Number RCS-102.
Each detector has 2 fission chambers which contain 14 grams U-235.
The total quantity is 14 chamber e ec rs) or 140 grams of U-235.
g d e i
3.5 Storate and Control of Material Incore and excore fission chambers will be stored in the Seabrook Station Source Room. Physical control of the detectors will be in accordance with the SNM inventory and control procedure as discussed in Subsection 1.4.2 of the application.
Materials other than special nuclear material specified in Section 3 will be controlled in accordance with Reference 5.
Reference 5 is the Radiation Protection Byproduct Material Program which is part of our NRC Byproduct Material License No. 28-20818-01.
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TABLE 1.1.4-1 Fuel Assembly Design
. Materials of Construction Component-Material Cladding Zircaloy-4 Grid assembly Inconel-718 cuide thimbles Zirealoy-4 Bottom nozzle 304-SS Top nozzle 304-SS Top nozzle springs and bolts Inconel-718 Core Mechanical Design Parameters Design RCC canless, 17 x 17 UO ' rods per assembly 264 2
Rod pitch (in.)
0.496 overall dimensions (in.)
8.426 x 8.426 Number of grids per assembly 8 - Type R Fuel Rods Outside diameter (in.)
0.374 Diametral gap (in.)
0.0065 Clad thickness (in.)
0.0225 Clad material Zircaloy-4 Fuel Pellets Material UO2 sintered Density (% of Theoretical) 95 Diameter (in.)
0.3225 Length (in.)
0.530
)
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TABLE 2.1.2 Type of Number Radiation Sensitivity Window Monitor, Survey.
Instrument Avail. Detected Range Thickness Etc.
2 Eberline E-520 2
Beta / Gamma 0.02 to 30 mg/cm Area surveys, Geiger Counter 2000 mR/hr posting boundaries with HP-270 External Probe Eberline BC-4 1
Beta O to 7 mg/cm2 Smear counting 1x106 counts 2
Personnel Eberline RM-l'4 2
Beta 100 to 7 mg/cm with HP-210 Probe 5x104 contamination cpm monitoring Eberline 1
Neutrons 0.5 to N/A Area surveys, PNR-4 5x103 mrem /hr-posting boundaries Eberline RO-2 2
Beta / Gamma 0.2 to 3.5 gm/cm2 Area surveys, 5000 mR/hr posting boundaries 2
Eberline RO-2A 2 ~
Beta / Gamma 2 to 3.5 gm/cm Area surveys, 5x104 mR/hr posting boundaries Johnson Extender 2
Gamma 0.1 mR/hr to N/A Area surveys, hot 1000 R/hr spots Eberline EC4-3 1
Gamma 1 to N/A Portable area with DAl-4 Probe 104 mR/hr monitor Eberline MS-2 1
Gamma 0 to 1/16" Al Smear, air sample Miniscaler 5x105 cpm counting Johnson Model 1
Beta 0 to N/A Tritium monitoring 955B Triton 1x104 Ci/m3 Ludium Model 12 2
Alpha 100 to 1 mg/cm2 Smear counting with Model 43-2 5x105 cpm Detector RADECO 4
N/A N/A N/A Part. air sampler 2
Part. air monitor Eberline AMS-3 1
Beta O to 1.4 mg/cm 1x105 cpm Surface Barrier 2
Alpha N/A N/A Alpha Spectroscopy
' Detector >
TABLE 2.1.2 (Continued)
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Type of Number Radiation Sensitivity Window Monitor, Survey, Instrument Avail.
Detected Range Thickness Ete'.
Beckman LS1800 1
Beta N/A N/A Tritium sample counting 2
Tennelec LB5100 1-Alpha / Beta N/A
.08 mg/cm Smears Canberra Series 1
Ganuna N/A N/A Isotopic analysis 85 of samples Lapel Air Sampler 1
N/A N/A N/A Air Sampler RADECO M809B2 2
N/A N/A N/A Particulate Air Battery Powered Sampler Air Sampler 1
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Item Dimension (Inches)
Canister Center-to-Center Spacing 10.35 i.060 i
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~ Boraflex Thickness.
0.071 i.010 Boraflex Length.
141,25 Boraflex Density
?.83 ga/cc Elemental Composition 416 Natural Boron Wrapper Thickness 0.02 Flux Trap Gap Thickness (Nominal) 1.086 f
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TABLE 2.2.8-1 Validation of KENO Methodoloxy with BNL Experiments Poison Plate Experiment No. - Enrichment Material eff q
5 - 2.35 w/o
.None 1.006 i 0.004 16 - 2.35 w/o Boral 1.006 1 0.004 31 - 4.30 w/o Boral 0.997 i 0.004 s
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Experiment Poison Data eff 0.9998 1 0.005 Core I Core XVI Boral Plate 0.9940 1 0.004 Core XIX Boral Plate 1.0046 1 0.004
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APPENDII A Experience and Traininst Records Health Physics Departinent Supervisor Health Physics Supervisors 4
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nr r r.o u t a n JOSEPH J.
RAFALOWSK1
~'
POSITION: Health Physics Department Supervisor
SUMMARY
OF QUALIFICATIONS:
Over eighteen years total nuclear power experience in the fields of radiation protection, health physics, primary and secondary chemistry, and nuclear plant mechanical equipment operation, maintenance and repair.
Eleven of these years have been in commercial nuclear power plants and over thirteen years.have involved various levels of supervisory experience.
Certified by the National Registry of Radiation Protection Technologist.
WORK EXPERIENCE:
Public Service Company of New Hampshire,Seabrook, New Hampshire 1978 to Present - Health Physics Department Supervisor (RPM)
. Responsible for the development implementation, and maintenance of the Station Radiological Safety Program. The scope of'the Station Radiological Safety Program provides for the planning, scheduling and directing of all radiological safety activities which ensure detection and control of radiation and radicactivity in the station and its systems to maintain personnel exposures As Low As Reasonably Achievable (ALARA).
Maine Yankee Atomic Power Company, Wiseasset, Maine 1971 to 1978 - Health Physics Technician Provided assistance in the development and implementation of the Radiological Safety Program. Responsible for assistance in continuous monitoring of all radiological activities within the plant and its systems, including the surrounding environment to ensure maximum safety at all t ime s.
Task assignments performed included, job coverage of radiological activities during outage and refueling periods involving high maintenance and reactor fuel movement, routine surveys for radiation, contamination and air sampling, calibration of portable and fixed radiation detection instrumentation, participation in radiological surveillance activities, participation in radiological emergen'cy planning activities including drills and exercises, authorization and establishment of controls for work to be ac'complished in radiological arets and supervision of assigned tech-nicians during outage and refueling periods.
The noted tasks assignments required, working in radiation and high radiation areas, low to very high contamination and airborne radioactivity areas, working with multiCurie sources of mixed corrosion products during decontamination and repair of components, monitoring for fission products during refueling periods and l
use of microcurie to multiCurie sources (alpha, beta, gamma and neutron) for calibration of radia tion detection instrumentation.
i Page 2 Joseph J. Rafalovski United Aircraft Rease~ arch Laboratories. East Hartford, Connecticut 1969 to 1971 - Senior Instrument Technician Responsible for radiological monitoring of work environment to ensure maxi-num sa fe ty a t all t ime s.
Performed research and development tasks ~
involving multicurie sources of Cobalt.60, Krypton 85, Iridium 192, Cesium 137 and Americium 241.
Calibrated check performed on radiation detection instrumentation using sealed sources up to millicurie range.
United States Navy 1965 to 1969 - Leading Engineering Laboratory Technician Responsible to ensure adherance to all radiological methods and procedures in accordance with Navy ships and Engineering Officer directives.
Performed radiological monitoring activities aboard nuclear submarines to ensure maximum safety at all times.
Calibrated radiation detection instru-mentation with sealed sources up to milliCurie range.
Provided job decontamination and repair of equipment. coverage involving multicuri Monitored for fission products during refueling period.
/
4
,i Joseph J. Rafalowski e
EDUCATION AND TRAINING 1963 to 1965 - United States Naval Nuclear Power and Engineering Laboratory Technician School, (15 months) 1976 - Basic Radiological Health Course, University of Lowell, Rowe, Mass.
(1 week) 1979 - Management and Supervision Courses (PSNH Sponsored), Seabrook, New Hampshire, (145 hours0.00168 days <br />0.0403 hours <br />2.397487e-4 weeks <br />5.51725e-5 months <br />) 1979 - Balance of Plant Course (Westinghouse), Seabrook, New Hampshire (60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br />) 1979 - Pressurized Water Reactor Information Course (Westinghouse), Seabrook, New Hampshire.
(60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br />) 1979 - EPRI-DOE Facility Decontamination Workshop, Hershey, Pa.
(2 days) 1980 - Planning for Nuclear Emergencies, Harvard School of Public Health, Boston, Mass.
(1 week) 1980 - Internal Radiation Dosime e.ry, University of Lowell, Lowell, Mass.
(1 week) 1980 - Assessment of Environmental Releases of Radioactivity,-University of Washington, Seattle, Washington (1. week) 1980 to 1981 - Continuing Education at Northern Essex Community College.
1981 - Applied ' Health Physics Course, Oak Ridge Associated Universities, Oak Ridge, Tenn.
(5 weeks) 1982 - Radiological Health Physics Course, University of Lowell, Lowell, Mass.
(100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />) l
William T. Cash
SUMMARY
OF QUALIFICATIONS:
Over eight years total work experience in the field of radiation protec-4 tion.
Seven of those years have been in commercial nuclear power.
Certified by the American Board of Health Physics.
WORK EXPERIENCE Public Service Company of New Hampshire,Seabrook, New Hampshire 1979 to Present - Health Physics Supervisor Responsible for the development of Station radiation protection policies programs, and specific procedures.
nicians in the performance of their duties and functions.I also supervise health ph Yankee Atomic Electric Company, Framingham, Massachusetts 1976 to 1979 - Radiation Protection Engineer Responsible for providing continuous technical assistance to engineerin operations, and plant staf fs on radiation protection matters.
g, Atomic operated three (3) nuclear power stations, Maine YankeeYankee Yankee, and Yankee Rowe.
, Vermont health ohysics group during' station outages.My responsibi to the plant the opportunity to work with both contained and non-contained radiThis afforded me material from non-detectable levels to multiCurie levels.
oactive tection controls were established for work in these environments wheRadiation pro radiation exposure rates could exceed 25R/hr of beta / gamma radiation re respiratory protection and protective clothing was required for work 'in
, where these environments, and special monitoring techniques were necessa detect neutron exposure, in'ternal contamina tion, and noble gas submersion ry to do se s.
University of Lowell, Lowell, Massachusetts 1975 to 1976 - Research Consultant Worked with non-contained radioactive sources formed by neutron acti in a research reactor.
vation for radioiodine.
I was analyzing different air sampling techniques i
I Page 2 William T. Cash Lovell Technological Institute (Nuclear Center), Lovell, Massachusetts 1972 to 1974 - Student Research Assistant I worked on a part time basis assisting the radiation protection personnel in routine survey operations.
ZDUCATION AND TRAININC 1970 to 1974 - Lowell Technological Institute Graduated 1974 Bachelor of Science in Radiological Health Physics 1974 to 1975 - University of Florida, Graduated 1975 Master of Science in Environmental Sciences with specialty in Radiological Health 1980 - American Industrial Hygiene Association, Akron, Ohio, " Respiratory Protection",
(1 week) 1981 - Health Physics Society Summer School, Lexington, Kentucky, " Reactor Health Physics",
(1 week)
+
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Stephen L. Dodge, III
SUMMARY
OF QUALIFICATIONS:
Nearly 10 years of radiation protection experience coupled with a B.S.
degree in radiological health physics.
engineering and auditing radiological control programs. Experience includes sup WORK EXPERIENCE Public Service Company of New Hampshire, Seabrook, New Hampshire 1979 to Present - Health Physics Supervisor Responsible to the Health Physics Department Supervisor for assisting in the development and implementation of Station policies, programs, proce-dures and work practices that collectively address all radiological con-siderations of nuclear power plant operation.
This includes supervision of department working foreman and technicians.
Power Authority of the State of New York, Buchanan, New York i977to1979-Assistant i
to the Radiological and Environmental Services Superintendent.
Responsibilities included providing technical and administrative assistance to the Radiological and Environmental Services Superintendent such as managing the plant's emergency preparedness program, administering the plant's whole body counting and personnel exposure records systems, and sies technicians providing radiological control function 500 radiation workers at one time.
The latter responsibility involved radiation () 20R/hr.), high surfacr contamination () authorizing a 1 x 106 cm.) and high airborne acivity.() hPCa) as well as handling significantdpm/100 sq.
quantities of radioactive ma terials.
i General Dynamics / Electric Boat Division, Croton, Connecticut l
1973 to 1977 - Radiological Engineer Responsibilities included managing temporary shielding installations in submarine reactor compartments to insure maintanance and testing tasks i
were performed in accordance with ALARA principles, performance of in-depth and surveillance type audits of the shipyard's radiological control program to insure compliance with applicable U. S. Navy and corporate requirements; and analyzing planned radiological work and shipyard business functions to recommend improvements in work practices, procedures and policie s.
I L
2
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Pcg3 2 Stephen L. Dodge III Brookhaven National Laboratory, Health and Safety Division, Upton, New York 1972 (Summer) - Student Research Assistant Responsibilities included calibrating radiation monitoring instrumentation with sealed sources, providing health physics coversge for research pro-jects and laboratory maintenance activities and performing radiation a d contamination surveys of laboratories, machine shops and accelerator equip-n ment and vaults.
EDUCATION AND TRAINING 1969 to 1973 - Lowell Technological Institute, Lowell, Massachusetts B.S. in Radiological Health Physics N
Rela ted Courses:
Radiation Safety and Control Nuclear Inst rumenta tion Radiation Biology Electronic Product Radia tion 1977
" Emergency Care for Radiation Injuries", New London Ct.,
(1 day) 1978
" Planning for Nuclear Emergencies" Harvard School of Public Health Boston, Massachusetts (5 days) 1978
" Radiological Emergency Response Planning" Defense Civil Preparedn Agency, Swan Lake, New York (5 days) ess 1980
" Radiological Health Physics Review",
Massachusetts (10 days)
Lowell University, Lowell, 1980
" Assessment of Environmental Releases of Radioactivity", University of Washington, Seattle, Washington (5 days) i.
i i
I.
t REFERENCE 1 4
Seabroo'k Station Unit 1 Special Nuclear Material License Application e
Contents - Excerpts from Section 9.1 of FSAR
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4 9.1.1.2 Facilities Description The new fuel storage facilities are located adjacent to the spent fuel pool in the fuel storage building to permit ease of handling of the new fuel into the transfer canal. The arrangement of the new fuel storage facilities is shown on Figures 1.2-15 through 1.2-21.
The storage vault is a rectangular concrete room containing the new fuel storage racks which securely hold the new fuel in a vertical position.
The storage racks are. individual vertical cells which are fastened together to form a module. All surfaces of the racks that come into contact with fuel assemblies are made of austenitic stainless steel, whereas the supporting structure is painted carbon steel.
The racks are constructed so that it is impossible to insert fuel assemblies anywhere in the storage vault except where holes are provided. The holes have a minimum center-to-center spacing of 21 inches in both directions which is sufficient to maintain the design margin of suberiticality, K gg 0.90, e
even if insnersed in unborated water.
New fuel assemblies are delivered to the' station in new fuel shipping con-tainers. These containers are off-loaded from the shipping vehicle and con-veyed into the fuel storage building where the fuel assemblies are removed, inspected and stored in the new fuel storage vault.
The new fuel is transported from the unloading zone to the storage vault and j
to the new fuel elevator by the 5-ton hook on the cask handling crane.
Security of new fuel is maintained by controlled access to the fuel storage building.
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Seabrook Station Unit 1 Special Nuclear Material License Application Contents - Excerpts from Section 9.1 of FSAR I
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s SB I&2 FSAR allows visual. observation of the insertion of drive rods into their proper locations in the vessel head, c.
Component Description 1.
Refueling Machine The refueling machine (Figure 9.1-3) is a rectilinear bridge and trolley system with a vertical mast extending down into the refueling water. The bridge spans the refueling cavity and runs on rails set into the edge of the refueling cavity.
The bridge and trolley motions are used to position the vertical mast over a fuel assembly in the core. A long tube with a pneumatic gripper on the end is lowered down out of the mast to grip the fuel assembly. The gripper tube is long enough so that the upper end is still contained in the mast when the gripper end contacts the fuel. A winch mounted on the trolley raises the gripper tube and fuel assembly up into the mast tube. The fuel is transported while inside the mast tube to its new position. All controls for the refueling machine are mounted on a console in the trolley.
The bridge is positioned on a coordinate system laid out on one rail, and a television monitor system on the console indicates the position of the brioge. The trolley is posi-tioned with the aid of a scale on the bridge structure which I
is read directly by the operator at the console. The drives for the bridge, trolley and winch are variable speed and include a separate inching control on the bridge and trolley.
The maximum speed for the bridge is 40 fpm; trolley and hoist speed is 20 fpm. The auxiliary monorail hoist on the refueling machine has a two step magnetic controller to give hoisting speeds of approximately 7 to 20 fpm.
Electrical interlocks and limit switches on the bridge and trolley drives prevent damage to the fuel assemblies. The winch is also provided with limit switches to prevent a fuel assembly from being raised above a safe shielding depth.
In an emergency, the bridge, the trolley and the winch can be operated manually using a hand-wheel on the motor shaft.
The refueling machine is designed to permit the handling of thimble plugs using a tool supported from the auxiliary hoist.
2.
Spent Fuel Pool Bridge and lloist The spent fuel pool bridge and hoist (Figures 1. 2-l fi, 1.2-18 and 1.2-21) is a wheel-mounted walkway, spanning the fuel storage area. which carries an electric monorail hoist on an overhead structure. The spent fuel pool bridge and hoist
(
9.1-18 j
SB 1&2 FSAR is used exclusively for handling fuel assemblies within the fuel storage area by means of a long handled tool suspended from the hoist. The hoist travel and tool length are designed to limit the maximum lif t of a fuel assembly to a safe shielding depth.
The spent fuel pool bridge and hoist has a two-step magnetic controller for both the bridge and hoist. The bridge speed is 11 and 33 fpm. Trolley travel speed is 3.5 fpm and the hoist lifting speed is 7 and 20 fpm. A hydraulic coupling is used in the bridge drive to limit starting acceleration.
A pushbutton pendant is provided for controlling bridge, trolley, and hoist motions. All pushbuttons are of the monen-tary contact type. Release of the pushbutton automatically stops motion and sets the brakes. Electrical interlocks ~
are provided to prevent damage to the fuel assemblies (see Subsection 9.1.4.3).
The spent fuel pool bridge and hoist is used to:
o Transfer new fuel from the new fuel elevator to the fuel transfer system container for passage into contain-ment through the fuel trans fer tube.
I o Transfer spent fuel from the fuel transfer system fuel container to the spent fuel storage racks.
o Transfer spent fuel from storage to the spent fuel shipping cask.
3.
Cask Handling Crane The cask handling crane (Figures 1.2-17, 1.2-18 and 1.2-20) is an electric overhead traveling crane with a main hook rated capacity of 125 tons and two 5-ton auxiliary hoists.
The bridge spans the new fuel storage area and the cask handling and decontamination areas. The crane serves the following functions:
o Upending new fuel containers and transferring new fuel to dry storage.
o Transferrin 6 new fuel from dry storage to the new fuel elevator (auxiliary hook).
o Trans ferring spent fuel shipping casks in and out of the cask loading and decontamination areas.
(
9.1-19 a
SB 1 & 2 Anendment 49 FSAR May 1983 i
The various speeds for the crane are as follows:
Bridge 30 to 50 fpm Main Hook Trolley 24 to 40 fpm Hoist 0.4 to 4 fpm Auxi,liary Hook (Monorail)
Trolley 50 fpm Hoist 3 to 30 fpm The drives for the bridge, trolleys, and hoists are variable speed with an inching control on the main hoist. Controls for all motions are full magnetic, 5 step, timed accelcration type. All motions can be controlled from either the operator's cab or pushbutton pendant controls.
4.
Polar Gantry Crane The polar gantry crane (Figures 1.2-5, 1.2-6 and RAI Figure 220.23-1) is an overhead gantry crane located in containment with a 103 foot diameter span to the rail centerline. The 49
(
main hoist has a rated capacity of 420 tons and the auxiliary hoist is rated at 50 tons. The various speeds (fpm) for the crane are as follows:
Minimum Maximum (Full Load)
(No Load) hain Hoist 0.2 3.5 (inching)
Auxiliary Hoist 1.9 30 (inching)
Bridge 40 50 Trolley 25 30 The polar crane is used during construction for installation of the reactor vessel and steam generators.
It is also used to lift the reactor lower internals as necessary during the life of the plant. The crane is used to remove and replace the reactor head and upper internals during refueling operations.
Magnetic controls provide variable speed for each crane motion.
The crane is arranged for cab and floor operation.
For a detailed discussion of the seismic analysis on the polar gantry crane, including the cable jerking effect, refer to the response to RAI 220.23.
49
(
5.
New Fuel Elevator The new fuel elevator (Figures 9.1-4 and 9.1-5) consists of a box-shaped elevator assembly with its top end open, and is sized to house one fuel assembly.
9.1-20
l The new fuel elevator is used exclusively to lower a new fuel assembly to the bottom of the fuel transfer canal where it t
is transported to the fuel transfer' system fuel container by the spent fuel pool bridge and hoist. The new fuel elevator hoist rated capacity is 3000 lbs with a lif ting speed of i
10 fpm. The hoist is provided with integral motor brake and load brake and a gear type limit switch with upper and l
lower limits. All portions of the elevator car which are immersed in water are stainless steel.
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6.
Fuel Transfer,, System The Fuel Transfer System (Figure 9.1-6, Sheets 1, 2 and 3) includes an underwater, electric-motor-driven, transfer car that runs on tracks extending from the refueling canal through t
the transfer tube and into the fuel storage building fuel t
transfer canal. A hydraulically-actuated lif ting arm is at each end of the trans fer tube. The fuel container in i
the refueling canal receives a fuel assembly in the vertical position from the refueling machine. The fuel assembly is then lowered to a horizontal position for passage through the trans fer tube. Af ter passing through the tube, the fuel assembly is raised to a vertical position for removal by
[
a tool suspended from the spent fuel pool bridge and hoist in the fuel storage area. The spent fuel pool bridge and hoist then moves to a storage loading position and places the spent fuel assembly in the spent fuel storage racks.
1 During reactor operation, the transfer car is stored'in the i
fuel storage area. A blind flange is bolted on the refueling I
canal end of the trans fer tube to seal the reactor containment.
The terminus of the tube outside the containment is closed by a valve.
I 7.
Rod Cluster Control Changing Fixture The rod cluster control changing fixture is a tool for changing rod cluster control elements in the reactor or the spent fuel pool (Figure 9.1-7).
The major subassemblies which comprise the changing fixture are the frame and track structure, the carriage, the guide tube, the gripper, and the drive mechanism.
The carriage is a moveable container supported by the frame and track structure. The tracks provide a guide for the four flanged carriage wheels and allow horizontal movement of the carriage during changing operation. The positioning stops on both the carriage and frame locate each of the three carriage compartments directly. below the guide tube. Two of these compartments are designed to hold individual fuel assemblies; the third is made to support a single rod cluster
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control element. Situated above the carriage and mounted 9.1-21
r SR 1 4 2 Amendment 49 FSAR gay 3983
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l on the refueling canal wall is the guide tube.
The guide tube provides for the guidance and proper orientation of the gripper and rod cluster control element as they are being raised and lowered. The gripper is a pneumatically actuated mechanism which engages the rod cluster control element.
It has two flexure fingers which can be inserted into the top of the rod cluster control element when air pressure is applied to the gripper piston. Normally the fingers are locked in a radially extended position. Mounted on the oper-ating deck is the drive mechanism assembly which is comprised of the' manual carriage drive mechanism, the revolving stop operating handle, the pneumatic selector valve for actuating the gripper piston, and the electric hoist for elevation control of the gripper.
8.
Spent Fuel Assembly Handling Tool _
i The spent fuel assembly han'dling tool (Figure 9.1-8) is used to handle new and spent fuel assemblies in the fuel storage area.
It is a manually-actuated tool, suspended from the spent fuel pool bridge and hoist, which uses four cam-actuated latching fingers to grip the underside of the fuel assembly top nozzle.
The operating handle to actuate the fingers is located at f
the top of the tool. When the fingers are latched, a pin is inserted into the operating hsndic which prevents the I
fingers from being accidently unlatched during fuel handling operations.
9.
New Fuel Assembly Handling Tool The new fuel assembly handling too'1 (Figure 9.1-9) is used to lif t and transfer fuel assemblies from the new fuel shipping containers to dry storage or to the new fuel elevator.
It is a manually-actuated tool, suspended from the cask handling crane which uses four cam-actuated latching fingers to grip the underside of the fuel assembly top nozzle. The operating handles to actuate the fingers are located on the side of the tool. When the fingers are latched, the safety screw is turned in to prevent the accidental unlatching of the fingers.
10.
Reactor Vessel Head Lift Rig The reactor vessel head lift rig (Fig. 9.1-22) consists of a welded and bolted structural steel frame with suitable rigging to enable the crane operator to lift the head and store it during refueling operations. The lift rig is permanently attached to the reactor vessel head. Attached to the head lift rig are the monorail and hoists for the reactor vessel
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L REFERENCE 3 Seabrook' Station Unit 1 special Nuclear Material License Application Contents - Excerpts from Section 13.4 of FSAR t
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SB 1&2 Amendment 47 FSAR September 1982 13.4 Review and Audit Operating phase activities that affect nuclear safety are reviewed and audited through a comprehensive program. The review and audit program will assure proper review and evaluation of proposed changes, tests, experiments, and unplanned events.
Regulatory Guide 1.33 and ANSI N18.7-1976/ANS 3.2 require-ments for reviews and audits will form the basis for the program.
13.4.1 On-Site Review 13.4.1.1 SORC A Station Operation Review Committee (SORC) will perform the on-site oper-ational review responsibilities. The function, composition, meeting frequency, responsibilities and authority of the SORC are contained in Technical Specification 6.5.1.
The goal of the SORC is to advise the Station manage-ment on all matters related to nuclear safety.
13.4.1.2 ' SORC Charter A SORC Charter, approved by the Station Manager and Vice President-Nuclear Production delineates the rules and procedures by which the SORC functions.
The Charter contains the following information:
47 Name Basis Purpose Authority Completion Meeting Frequency Quorum Committee Responsibilities Records Amendments Endoresement The SORC composition and qualifications are provided in Technical Specification
- 6. 5.1..
The qualification levels of Station staff personnel and their alternages assigned to SORC membership meet or exceed those required by Section 4 of ANSI /ANS 3.1-1978.
In performing its duties the SORC will establish its own rules of practice that include:
When less than full membership is present, the quorum will ensure a.
that matters to be considered are limited to those that are within the technical competence of the members present, j
4fI l
13.4-1 L
r SB 1 & 2 Amendment 53 FSAR August 1984 b.
Committee members will ensure an appropriate interdisciplinary review of activities under discussion.
c.
Sub-committees may be used at the discretion of the Chairman.
When used, due consideration shall be given to the interdisciplinary composition of the subcommittee membership.
d.
The Chairman may authorize the use of experience from sources outside the SORC or outside the station staff where the particular matters under consideration cannot otherwise be reasonably resolved.
The minutes of each SORC meeting are official plant records and e.
shall be retained as provided in the station record retention pro-cedures.
f.
In addition to distribution of SORC meeting minutes as provided in the Technical Specification 6.5.1, copies will be submitted to other appropriate management.
13.4.1.3 Operations Phase Reviews The scope of SORC review matters include those noted in Section 4.4 of ANSI 18.7-1976/ANS 3.2 as endorsed by Reg. Guide 1.33, The general topics to be addressed include:
Station procedures and changes thereto that affect nuclear safety.
a.
b.
Proposed changes to the Operating License or Appendix A Technical Specifications.
Proposed modifications to nuclear safety-related structures, c.
systems, and components.
d.
Proposed tests and experiments which affect nuclear safety and are not addressed in the FSAR or Technical Specifications.
Evaluation of uplanned events that affect nuclear safety.
e.
f.
Review of events requiring 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> written notification to the Nuclear Regulatery Commission.
13.4.1.4 S tart-up PhIse Reviews The SORC will be activated a minimum of six months prior to fuel load in order to conduct the following activities:
Review test results of the integrated system pre-operational tests l
a.
performed prior to fuel load of each unit.
53 1
45 13.4-2
i SB 1 & 2 Amendment 53 l
FSAR August 1984 2.
Maintain surveillance of plant operations and maintenance activities to provide independent verification that these activities are performed correctly and that human errors are reduced as far as practicable.
3.
Perform independent reviews and audits of plant activities including maintenance, modifications, operational problems, and operational analysis, and aid in the establishment of programmatic requirements for plant activities.
4.
Where useful improvements can be achieved, this group will develop and present detailed recommendations to corporate for such things as revised procedures or equipment management modifications.
b.
The OES is not re spons ib le for sign-off functions such that it l
becomes involved in the operating ' organization.
53 -
13.4.3.2 Reports The OES will prepare written summaries of reviews and evaluations performed These sunmaries will include the results of, 'and recommendations l
as noted above.
resulting from, SS such reviews and evaluations.
Monthly reports containing a aummary of work completed and recommendations made will be forwarded to the Training Center Manager, with information copies to the Vice President -
Nuclear Production and NHY Senior Vice President.
13.4.3.3 Charter 45 46 55 The composition, qualifications, duties and responsibilities, and reporting requirements stated above will be incorporated into the OES Charter.
4s55 13.4-5 l
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l REFERENCE 4 1
Seabrook Station Unit 1 Special Nuclear Material License Application Contents - Excerpts from Section 1.8 of FSAR I
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Regulatory Guide 1.5 Assumptions Used for Evaluating the (Rev. O, 3/71)
Potential Radiological Consequence of a Steam Line Break Accident for Boiling Water Reactors This regulatory guide is not applicable to Seabrook Station.
Regulatory Guide 1.6 Independence Between Redundant Standby (Rev. O, 3/71)
(Onsite) Power Sources and Between Their Distribution Systems The design totally conforms with the recommendations of this regulatory guide.
The subject matter of this guide is discussed in Subsections 8.3.1 and 8.3.2.
Regulatory Guide 1.7 Control of Combustible Gas Concentrations (Rev. 2, 11/78) in Containment Following a Loss-of-Coolant l
Accident 44 The Seabrook Station employs a large dry containment for containing fission gases and aerosols following an accident, in accordance with CDC 50.
Any hydrogen generated during an accident is controlled per GDC 41.
Regulatory Guide 1.7, Rev. 2 details an acceptable method of showing compliance with CDC.
The design of the Seabrook plant, considering the Westinghouse
/
k scope of supply, was analyzed against CDC 41 and 50, using assumptions and models specified in Rev. 2 of this guide.
44 The BOP design complies fully with Regulatory Guide 1.7, Rev. 2.
The amount of Zr-H2 reaction used for evaluation of combustible gas control is 1.5%.
This is based upon five times the calculated amount of Zr reacting from the ECCS performance analysis, as permitted by the regulatory guide, and was used in lieu of the blanket 5%.
Refer to Subsection 6.2.5 for further discussion of this subject.
Regulatory Guide 1.8 Personnel Selection and Training (Rev. 1-R, 9/75; reissued 5/77)
Endorses ANS 3.1/ ANSI N18.1-1971 l
44 The personnel selection and training program meets the requirements of Regula-tory Guide 1.8 (1977 edition), except that ANSI /ANS 3.1-1978 will be used as the standard rather than ANS 3.1/ ANSI N18.1-1971.
44Property "ANSI code" (as page type) with input value "ANSI N18.1-1971.</br></br>44" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process. 1.8-3 L
l SB 1 & 2 Amendant 52 FSAR December 1983 With regard to Section 4 of ANS 3.1-1978, individuals who do not possess the
(
formal requirements specified in this section shall not be automatically eliminated where other factors provide sufficient demonstrations of their abilities.
For further clarifications and alternatives, see discussion in Sections 13.1, 13.2,16.6.3,16.6.4 and 17.2, and in Reg. Guide 1.58 appearing later in this section.
44 Regulatory Guide 1.9 Selection of Diesel Generator Set (Rev. 2, 12/79)
Capacity for Standby Power Supplies l
44 The diesel generator design is in general conformance with the recommendations of Regulatory Guide 1.9.
l 44 Position C.14 requires that the engine run at full load for 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br />, following 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> at short-time rated load. For Seabrook, the " Load Capability Qualifi-cation" test was performed as per IEEE 387-1977. The engine was run at full load for 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> after reaching equilibrium temperature, followed by 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> at the short-time rated load.
The subject matter of this guide is discussed in Subsections 8.1.5 and 8.3.1.
44 Regulatory Guide 1.10 Mechanical (Cadweld) Splices in Reinforcing (Rev. 1, 1/73)
Bars of Category I Concrete Structures The requirements for crew qualification, inspection, testing and sampling of mechanical splices in reinforcing bars' of concrete structures comply with the regulatory positions outlined in Regulatory Guide 1.10, Rev. 1, except that retesting due to the failure of production and sister mee.hanical splices is in accordance with ASPE Section III, Division 2,1975 edition and Article CC-4333.4.5(b) of the Winter 1979 Addenda in lieu of the corresponding article in the 1975 code. For further discussion on this subject, refer to Subsection 3.8.1.6.
g This regulatory guide was withdrawn on July 8, 1981, and superseded by Reg.
Guide 1.136, Rev. 2, 6/81 Regulatory Guide 1.11 Instrument Lines Penetrating Primary (Rev. O, 3/71)
Reactor Containment The design of instrument lines penetrating primary containment conforms with the intent of Regulatory Guide 1.11, with the exception of the containment pressure monitoring lines. For these lines, isolation from the containment atmosphere is provided by a sealed bellows arrangement located immediately adjacent to the outside containment wall, and connected to the pressure trans-citter outside containment by a sealed fluid tube.
Isolation outside contain-ment is provided by the diaphragm in the pressure transmitter. The justifi-cation for this special arrangement results from the importance of these connections to sense accident conditions and initiate safeguard actions.
1.8-4 L
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REFERENCE 5 Seabrook Station Unit 1 Special Nuclear Material License Application Contents - Radiation Protection Byproduct Material Program Excerpt from Byproduct Material License No. 28-20818-01 I
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f RADIATION PROTECTION BYPRODUCT MATERIAL PROGRAM I
1.0 In trod uc tion The Seabrook Station Radiation Protection Program (RPP) addresses the rules and regulations established by the Nuclear Regulatory Commission (NRC) for t
the control of byproduc t radioactive ma terial. The program was developed to protect the public health and safety and to ensure radiation expos,ures j
to station employees and contractors are maintained as low as reasonably achievable. This program' applies to radioactive material that is used or s tored at Seabrook Sta tion.
The program establishes requirements concerning radiological surveys and postings, radioac tive ma terial procurement, receipt and use, radioa c tive material control, contamination control, personnel training, exposure moni-l toring and dosimetry, ins trument calibration, radioactive source inventory and leak testing, radioactive vas te disposal, record keeping requirements,
i and notifica tion of incidents.
3 The program is approved for use by the station manager and vice president of nuclear production following concurrence by the health physics depart-ment supervisor.
The program shall remain ef fective during the period that a NRC Radioactive i
Byproduct Material License is maintained for Seabrook Station.
Amendmen ts to and renewal of the license will require review of the program by company management and the Radiation Safety Committee to ensure adequate controls are maintained.
j j
2.0 Radiation Protection Organization f
2.1 Radiation Safety Committee The program establishes a Station Radiation Safety Committee (RSC) charged with the responsbiliity of administra ting, interpreting, and enforcing sound radiological protection practices in conformance with NRC regulations and industry standards. The RSC is ultimately respon-sible for providing the adminis tra tive controls, operational proce--
j dures and management review necessary to assure safe operation of the l
- program, t
3 Membership on the committee consists of the station manager, the
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health physics department supervisor, the chemistry department i
supervisor, health physics supervisor, and an I&C department engineer.
i Designated alternates are specified for each p6sition.
The station manager sets as committee chairman with the health physics department supervisor acting as chairman in the absence of the station manager.
The RSC meets periodically as necessary, but at least quarterly.
A minimum of three members is necessary to meet the quorum.
Committee decisions, authorizations and other matters of record are documented i
in mee ting minutes.
l i
2.2 Health Physics Department Organization The health physics department supervisor is assigned the day to day administrative authority and responsibility for control and super-vision of licensed byproduct material.
He is designated as the radiation safety officer for Seabrook Station. He has the adminis tra-tive authority and responsibility to review, audit and approve the procedures, locations, and users of licensed byproduct material.
He delegates, in his absence, this authority and responsibility to a health physics supervisor.
The health physics supervisors report to the health physics department supervisor. The HP supervisors are responsible for supervising the day to day operations of the program including performance of surveys, generation of records, cantrol of licensed radioactive material, and dosimetry issue.
Only individuals authorized to use licensed byproduct material will be allowed to do so.
Authorization is granted when there is need to use radioactive ma terial and when users are trained in safe use and handling.
3.0 Personnel Training Personnel working with radioactive ma terial are trained in the proper use and handling commensurate with their work responsibilities. The objective of the training is to provide information - to help minimize a person's expo-in keeping with ALARA principles and to prevent misuse of sources.
sure Topics covered in the training will include as necessary:
- Safe handling of radioactive ' ma terials.
- Techniques of radiation and contamination surveys,
- Use and calibration requirements of radiation detection instruments,
- Pos ting requirements of radiation areas, Exposure control, Dosimetry devices, As low as reasonably achievable (ALARA) considerations, Biological ef fects of ionizing radiation, and
- Response to emergency events.
Training may be conducted by health physics department personnel or the health physics training instructor. Retraining is conducted on an annual basis for those individuals performing activities under this program.
Training vill be documented.
Individuals with the need to handle or use radios'ctive ma terials without documented training may do so under the direct supervision of 'someone who has been trained.
c
4.0.Procuremen t. ' Receipt and Handling of Radioactive Material 4.1 General Users of byproduct radioactive material notify health physics prior to procurement of non-exempt sources in order to ensure appropriate preparations are made prior to source arrival.
For radiation protection purposes, health physics maintains control over radioactive material storage areas. A designated storage area may be a storage cabine t, safe, or entire room, depending on the radiological and physical constraints associated with the radiological material. Non-exempt quantities of byproduct material are only stored in locations approved by health physics department supervisor or his designee. Sources which are an integral part of plant equipment (i.e., radioactive monitors) may be considered stored by virtue of their mounting in the device / equipment.
4.2 Receipt Radiological surveys are performed upon receipt of non-exempt radio-active material to verify radiological conditions and integrity of packaging in accordance with 10CFR 20.205.
Receipt of non-exempt sources is logged into a source inventory and control system.
4.3 Control of Source Storage Areas Radiological surveys are performed of source storage areas upon ini-tial storage of non-exempt sources or when changing storage locations if radiological conditions could change as a result of the source relocation. Movement of non-exempt sources outside of storage loca-tions is documented in the inventory and control system.
Persons requesting use of sources must be authorized as discussed in RPP Sec tion 2.2.
Storage areas are posted in accordance with 10CFR 20.203 and Section 6.0 of this program manual.
4.4 Handling of Radioactive Material Byproduct radioactive material is handled in such a way as to ma in-tain radiation exposure as low as reasonably achievable.
Prior to initial use of licensed radioactive material, a safety evaluation is conducted by a representative from the user department and a health physics representa tive. The evaluation should address the following considerations with respect to the particular radioactive materials adequacy of facilities and equipment, training and experience of users and opera ting or handling procedures.
The evaluation is submitted to the RSC for review, comment and approval.
If necessary, a controlled and monitored fume hood 'is available for use with volatile or high levels.of uncontained radioactive ma terial.
Radioactive byproduct ma terial is handled in such a way as to pre-vent exposure to individuals in uncontrolled areas to radiation levels as defined in 10CFR 20.105.
4.5 Source Inventory and Leak Checks Periodic inventories and leak checks are performed to account for all licensed sealed sources and to verify that leakage has not occurred.
9 The radioactive licensed sealed source inventory is peformed quarterly.
Each licensed sealed source (other than hydrogen-3), with a half-Life greater than 30 days, and in any form other than gas, is tested for leakage and/or contamination at intervals not to exceed six months.
Leak check of sealed sources is not required if material is in storage and not in use.
However, a leak check is required prior to.its move-ment froe storage if one has not been performed within six months..In addition, a leak check prior to transfer off-site is required.
The leak test is capable of detecting the presence of 0.005 micro-curies of radioactive ma terial on the test sample. This te s t sample is taken from the surfaces of the sealed source, or when permanently mounted or stored, where contamination might be expected to accumu-late.
Records of leak test results are kept and maintained for inspection.
If the leak test reveals the presence of.005 uCi or more of loose con tamina tion, the source shall be immediately withdrawn from use and handled according to the instructions of health physics supervisory personnel. A report shall be filed within 30 days of the leak test with the director of the Nuclear Regulatory Commission Inspection and Enforcement Regional Office. The report shall include the equipment involved, the test results and corrective action taken.
5.0 Radiological Surveillance Routine surveys are performed on'a periodic basis.
Other surveys may be performed at the discretion of health physics persont.el to ensure the control of radioactive material and/or personnel exposure.
Radiological surveys are performed to ensure adequate controls are main-tained over radioactive ma terials, to es tablish and maintain pos ted radiation boundaries and to control radiation exposure within these posted areas.
Contamination surveys are performed to ensure tha t ' radioactive ma terials are properly contained and have not spread to other areas.
Non-routine radiation surveys and airborne surveys-are performed at the discretion of health physics personnel.
Events that may require these surveys include a dropped source, spill, suspected contamination, and waste shipments.
Surveys specifically required by procedure are documented.
Surveys per-formed for informational purposes need not be documented.
5.1 Radiation Surveys Area radiation surveys are performed by health physics personnel in those areas in which byproduct material is used and stored, (e.g.,
calibration facility).
Appropriate calibrated survey ins truments are used for beta-gamma dose ra te measurements.
3 i
Bioassay services are provided whenever an internal intake of radioactive material is known or suspected to have occurred.
Due to the nature and j
amount of byproduct material expected on site prior to plant opera tions, i
the need for a routine bioassay program is not expected.
New Hampshire Yankee may, provide in vivo bioassay (whole body counting) either on site by Seabrook Health Physics Department or through contracted services.
If j
required, additional bioassay services are provided through contractor ser-vices with Yankee Atomic Electric Company, Framingham, MA.
Solid waste generated during decontamination efforts is begged, sealed and labeled as radioactive material and disposed of in accordance with Section 10 of this program.
1 Respiratory protective equipment is available, although it is not antici-pated that it will be required prior to plant operation. However, gloves, i
coveralls, shoe covers, head covers, laboratory coats and handling tongs will be available for those handling unsealed byproduct material if the need arises. The use of unsealed ma terial in a manner that might require protective clothing, is done in accordance with proper radiation protection j
procedures as outlined in Section 4.4 of this program.
I 8.0 Exposure Monitoring Personnel exposure monitoring is performed for each individual as required i
by 10CFR 20.202.
The need to wear dosimetry devices is determined by a health physics supervisor through evaluation of the potential for exposure on a case by case basis.
Thermoluminescent dosimeters (TLDs) and/or self reading pocket dosime ters (SRPDs) are used to monitor personnel exposure to ionizing radia tion.
The i
dosimetry is provided by an on site dosimetry system operated by health physics. This system may be supplemented by contracted services if required. The TLDs are normally changed out on a monthly basis.
SRPDs are normally read out at the end of each day for exposure tracking purposes.
- fhe SRPDs are calibra ted periodically (i.e., six month intervals).
They are leak tes ted to check tha t leakage current is less than 5% of full scale af ter 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> and accuracy is checked to within + 10% of known values.
Previous occupational exposure records (as required by 10CFR 20.102) and current occupational exposure records are maintained by health physics 4
department. Current exposure records are normally updated monthly but.not less than quarterly in accordance with 10CFR 20.401.
J Reports of exposure to individuals will be generated when reques ted by the i
employee in accordance with 10CFR 19.13, when required by 10CFR 20.409 for worker termination and reports to the Commission.
9.0 Instrumentation and Calibra tion Radiation detection ins trumentation is used at the sta tion to ensure that all provisions of the byproduct material license are met, and to ensure that personnel exposure to ionizing radiation is as low as reasonably i
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5.2 Contamination Surveys Loose surf ace contamina tion is quantified by using a towel wire or a disc " smear" on a surface to be s u rvey e d.
For large areas a represen-tative area surveyed per smear is typically 100 sq cm.
Surface con-tamination surveys are performed in those areas in which rasealed sources are used or atored (eg. chemis try primary labs, FP counting rooms). The smears and/or wipes are analyzed with counting room instruments or portable count rate survey maters whichever is mos t appropriate.
Fixed contamination surveys may be performed when surface contamina-tion has been worked into surfaces and is no longer loose.
Portable survey instruments are used. These surveys are performed at the discretion of the HP department supervisor.
5.3 Airborne Surveys Air sampling is conducted for particulate radioisotopes, radioiodine, noble gases, and tritium, as appropriate, whenever circums tances create the potential for airborne radioactive material.
Par ticula te, radiciodine and noble gas samples are analyzed and quan-tified with gross counting techniques or gamma spectroscopy as warranted by specific circums tances. Tritium samples are analyzed and quantified using a liquid scintillation counter.
6.0 Es tablishing and Posting Radiologically Controlled Areas Radiologically controlled areas are established to control access and to warn personnel of the hazards of ionizing radiation. The types or radiolo-gically controlled areas that may be es tablished are Radia tion Area, High Radia tion Area, Radioactive Material Area, and Airborne Radioactivity Area.
These areas will be posted, barricaded and controlled in accordance with 10CFR 20.203.
Surveys for establishing, updating or removing postings of radiologically controlled areas are performed as required.
7.0 Contamination Control Contamination is unsealed radioactive ma terial located in an area where it is unwanted.
Controls are established to monitor for its presence and pre-vent or minimize its spread or occurence.
Should an individual.. item or area become contaminated above the limi ts indica ted below, actions are taken to remove the contamination for control access. For radioactive sealed sources, measures are taken as outlined in Section 4,5 of this manual.
Specific loose surface contamination control limits are as follows:
-Radioactive sealed sources:
.005 uti
-Personnel, personal items, 1000 dpm/100 sq cm beta-gamma equipment and material:
20 dpm/100 sq cm alpha
-Areas:
1000 dpm/100 sq cm beta-gamma 100 dpm/100 sq cm alpha
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achievable (ALARA).
The s ta tion maintains ins trumenta tion capable of measuring all anticipated exposure levels during normal source handling operations and during abnormal occurrences.
Adequate quantities of ins tru-ments are available to ensure operating instruments at all times.
The instrumenation is also capable of detecting beta-gamma and alpha con-tamination. levels at leas t as low as those specified in Section 7 of this program. Sampling devices are maintained that allow the s tation health physics staff to monitor for airborne activity.
A computer based gamma spectroscopy system is maintained on site to provide accurate qualitative and quantita tive radioisotope analysis.
Calibration and response checks of radiation detection instruments are per-formed to assure continued, accurate response. ' Ins truments are calibra ted with sources or me thods traceable to the National Bureau of Standards on a periodic basis (i.e., semi-annually) cr af ter any repairs that could af fect instrument response. Dose ra te and count ra te instrumentation are adjusted to respond within + 10% of the true value and are adjusted at the points on each scale according to ANSI N323, Radiation Protection Ins trumenta tion Test and Calibration.
Efficiencies for scalers and other contamina tion measuring instruments are determined to allow accurate conversion of observed count ra tes to activity.
Daily. tns trumen t checks are performed on scaler and other contamination measuring laboratory ins truments when in use.
These functional checks are recorded on health physics department forms.
HP personnel are qualified in the use of radiation detection instruments by training and/or experience. Records associa ted with the maintenance, calibration and use of HP instruments are maintained by the health physics department.
10.0 Vaste Disposal Disposal of byproduct ma terial was te prior to station opera tion is not anticipated.
In the event tha t small quantities of radioactive vas te are generated as a result of byproduct material use, the waste will be placed in a safe, strong, tight container and stored in an approved radioactive material storage area.
If quantities neceseica te disposal prior to s ta tion operation, arrangements will be made with a commercia.' waste disposal ser-vice, and the waste will be shipped in accordance with t'ederal regulstions 10CFR 61, 10CFR 71, and 49CFR. Licensed sealed sources tha t are found defective shall be removed from service and e ither disposed of in a safe container and stored in an approved storage loca tion or re turned to the manufacturer for repair.
By policy, Seabrook Station will not normally dispose of radioactive by-product material by release into sanitary sewerage systems or other uncontrolled areas except as specifically approved by the Radia tion Safe ty Committee.
l i
l 11.0 Notificasion Requirements
]
For the purposes of the Byproduct Material License, possible events tha t could require initiation of emergency actions and NRC notifica tion-are the following:
Loss or thef t of a radioactive source
- Leakage of radioactive material from a sealed source Overexposure of personnel Excessive concentrations of radioactive material Most s i tua tions that could be encountered can be handled with minimal expo-sure or spread of contamination. The Radiation Protection Program stresses the prevention of the above lis ted events.
Notification of such events will be performed in accordance with 10CFR 20.402, and 10CFR 20.403.
Immedia te notifica tion will be made when necessary to the adminis tra tor, NRC Regional Of fice.
r 1
1 P
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4
I AHaamed A Shu_41 4 2-SB 1 & 2 Amendment 53 FSAR August 1984 the assembly upon attempted withdrawal.
New fuel rack design also requires on that the de formation of the impac ted storage cells not adversely af fec t the minimum spacing requirements of 21 inches.
Provisions have been made in the crane handling system, by providing load limit switches, to insure that the maximum uplift force specified for the design of new fuel rack is not exceeded, thus averting any increase in K gg.
e Protection of the new fuel storage facilities from wind and tornado ef fects D
is discussed in Section 3.3.
Flood protection is discussed in Section 3.4.
~
Missile protection is discussed in Section 3.5.
Protection against fire hazards is discussed in Section 9.5.1.3.
Radiation monitoring is provided to meet the requirements of 10CFR50, Appendix A, GDC 63;-=d 10CF"J0.24.
24 The radiation monitor is a GM tube based area monitoring channel. An alarm is initiated in the control room when the radiation level exceeds a pre-determined setpoint (see Table 12.3-13).
Details of the radiation monitoring system are provided in Section 12.3.4.
9.1.1.2 Facilities Description The new fuel storage facilities are located adjacent to the spent fuel pool in the fuel storage building to permit ease of handling of the new fuel into the trans fer canal. The arrangement of the new fuel storage facilities is
- shown on Figures 1.2-15 through 1.2-21.
(
The storage vault is a rectangular concrete room containing the new fuel storage racks which securely hold the new fuel in a vertical position.
The storage racks are individual vertical cells which are fastened together to form a module. All surfaces of the racks that come into contact with fuel assemblies are made of austenitic stainless steel, whereas the supporting structure is painted carbon steel.
The racks are constructed so that it is impossible to insert fuel assemblies anywhere in the storage vault except where holes are provided. The holes have a minimum center-to-center spacing of 21 inches in both directions which is suf ficient to maintain the design margin of suberiticality, K gg 0.90, e
even if immersed in unborated water.
New fuel assemblies are delivered to the station in new fuel shipping con-tainers. These containers are off-loaded from the shipping vehicle and con-veyed into the fuel storage building where the fuel assemblies are removed, inspected and stored in the new fuel storage vault.
The new fuel is transported from~ the unloading zone to the storage vault and to the new fuel elevator by the 5-ton hook on the cask handling crane.
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Security of new fuel is maintained by controlled access to the fuel storage building.
9.1-2 l
l l
Attaehmen+ A
$hech 2 of 2.
SB 1 & 2 Amendment 52 FSAR December 1983 2.
The detectors provide on-scale readings of dose rate that include the design maximum dose rate of the radiation zone in which they are located, as. well as the maximum dose rate for anticipated operational occurrences.
i j-3.
Each monitor has local visual and audible alarms, with variable i
set point.
i 4.
Indication and annunciation is available in the main control room.
5.
The design objectives and location criteria are in conformance with 10CFR Part 20, "::: ?^ er.2 Part 50, Appendix A,' General Design Criteria 63 and 64, and Regulatory Guides 1.21, 8.2.
and 8.8.
i 6.
Post-accident monitoring instrumentation is provided as dis-cussed in Section 7.5.
b.
System Description
The digital computer-based radiation data management system (RDNS) 4 consists of local microprocessors for each channel, interconnected
[
by a redundant communication loop to a redundant host computer j
system. The host computer system is common to both Unit 1 and Unit 2 radiation monitoring channels. Either of the two computers can by itself provide the total computing capacity required for satisfactory ' operation of the RDMS for both Units 1 and 2.
The host computer system, in turn, is connected to an operator dis-i play / control console in each unit control room and the health i
physics control point. The area radiation monitoring system i
instrument engineering diagram, Figure 12.3-18, shows an overview 61 1-of the system, its components and location.
i Table 12.3-13 lists the various area radiation monitoring channels provided and their pertinent design informat'lon, such as detector-type, range, background radiation, safety class, alarm setpoints, referenced drawings for location of area radiation detectors, etc.
The components for the area radiation monitoring system including the computers, are supplied from diesel-backed busses, except for j
Class 1E area radiation equipment that is supplied from Class 1E uninterruptible power supplies (UPS).
Except for the post-LOCA containment monitors and other high range -
monitors, each channel is equipped with a radioactive check source j
which can be actuated from the main control room during test. The post-LOCA monitors and other high range monitors use an electronic signal to test the circuit.
12.3-15'
+
..